ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 121

Document Title:

Nevada Test Site – Occupational External Dose

Document Number: ORAUT-TKBS-0008-6 

Revision: 01

Effective Date: 07/30/2007

Type of Document TBD

Supersedes: Revision 00

Subject Expert(s): Richard V. Griffith

Site Expert(s): N/A

Approval: Signature on File Approval Date: 07/01/2007

Eugene M. Rollins, Document Owner

Concurrence: Signature on File Concurrence Date: 07/03/2007

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 07/02/2007

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 07/27/2007

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 07/30/2007

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 2 of 121

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

09/21/2004 00 Information contained in Revision 00 is applicable only to employment 

periods after 1962 (post-atmospheric testing phase) and to workers not

1. BLANCA (October 30, 1958)

identified as involved with drillback activities prior to 1965. In addition, 

Revision 00 is not applicable to dose reconstruction for (1) workers 

involved with weapons testing at locations other than the NTS (South 

Pacific, Alaska, etc.); (2) workers affected by any of the following 10 

underground tests that resulted in unexpected release of radioactive 

material:

2. DES MOINES (June 13, 1962)

3. BANEBERRY (December 18, 1970)

4. CAMPHOR (June 29, 1971)

5. DIAGONAL LINE (November 24, 1971)

6. RIOLA (September 25, 1980)

7. AGRINI (March 31, 1984)

8. MIDAS MYTH (February 15, 1984)

9. MISTY RAIN (April 6, 1985)

10. MIGHTY OAK (April 10, 1986) 

It is also assumed that, because they were such a rarity, it would be 

likely that if a claimant was involved in one of these events, he [or his 

survivor] would mention it in his CATI. All of these conditions are likely 

verifiable by careful examination of the DOL, DOE (incident reports 

and dosimetry records), and OCAS documents. Incorporates internal 

and NIOSH comments. Initiated by Eugene M. Rollins.

07/21/2006 00 PC-1 Approved page change revision incorporates definitions and directions 

for dose reconstruction for non-presumptive cancers that are excluded 

from the 1951 through 1962 Special Exposure Cohort. Text was 

added or modified on pages 8-9 in Section 6.1. Incorporates NIOSH 

formal review comments. No sections were deleted. This revision 

results in no change to the assigned dose and no PER is required. 

Training required: As determined by the Task Manager. Initiated by 

Eugene M. Rollins. Approval:

Eugene M. Rollins, TBD Team Leader

Signature on File 07/12/2006

John M. Byrne, Task 3 Manager

Signature on File 07/12/2006

Edward F. Maher, Task 5 Manager

Signature on File 07/14/2006

Kate Kimpan, Project Director

Signature on File 07/20/2006

James W. Neton, Associate Director for Science

Signature on File 07/21/2006
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 3 of 121

01/11/2007 00 PC-2 Approved page change revision incorporates coworker data for 

completion of external dose prior to universal badging in April 1957. 

Text was added or modified on pages 40-42 in Sections 6.3.4.1. 

Added DeMarre reference on page 50. Deleted the term, claimant 

favorable, from the Glossary on page 54. As a result of changes, 

corrected table references on page 46 and 48. No further changes 

occurred as a result of internal formal review. Incorporates NIOSH 

formal review comments. No sections were deleted. This revision 

results in an increase in assigned dose and a PER is required. 

Training required: As determine by the Task Manager. Initiated by 

Eugene M. Rollins. Approval:

Eugene M. Rollins, TBD Team Leader

Signature on File________________________________12/15/2006

John M. Byrne, Task 3 Manager

Signature on File________________________________12/12/2006

Edward F. Maher, Task 5 Manager

Signature on File________________________________12/12/2006

Kate Kimpan, Project Director

Signature on File________________________________01/11/2007

James W. Neton, Associate Director for Science

Brant A. Ulsh Signature on File for__________________ 01/11/2007

07/30/2007 01 Approved Revision 01 initiated to add Section 6.4 Dose Reconstruction 

Recommendations, Section 6.5 Claim Analysis Methods. Added 

Attachments A through D. Constitutes a total rewrite of the document. 

Incorporates formal internal and NIOSH review comments. Adds 

Attributions and Annotations section. Training required: As determine 

by the Task Manager. Initiated by Eugene M. Rollins. 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 4 of 121

TABLE OF CONTENTS

SECTION TITLE

Acronyms and Abbreviations ...............................................................................................................

PAGE

8

6.1 Introduction ............................................................................................................................11

6.1.1 Purpose......................................................................................................................12

6.1.2 Scope.........................................................................................................................12

6.2 Dosimetric Basis of Comparison ............................................................................................ 12

6.2.1 Photon Measurement Quantities ................................................................................13

6.2.2 Beta Measurement Quantities ....................................................................................14

6.2.3 Neutron Measurement Quantities ...............................................................................14

6.3 Dose Reconstruction Parameters........................................................................................... 14

6.3.1 Administrative Practices .............................................................................................18

6.3.1.1 Exposure Limits ........................................................................................... 20

6.3.1.2 Dosimeter Exchange.................................................................................... 20

6.3.1.3 Dosimetry Codes for External Monitoring ..................................................... 21

6.3.1.4 Duplicate Dosimetry Records....................................................................... 22

6.3.1.5 Special Issues.............................................................................................. 24

6.3.1.5.1 External Exposure to “Hot Particles” ........................................... 24

6.3.1.5.2 Intentional Nonuse of Dosimeters ............................................... 25

6.3.2 Individual Monitoring Methods ....................................................................................25

6.3.2.1 Beta/Gamma Dosimeters ............................................................................. 25

6.3.2.1.1 Beta Dosimetry with Film Badges ............................................... 28

6.3.2.2 Neutron Dosimeters ..................................................................................... 29

6.3.2.2.1 Film Dosimeters.......................................................................... 29

6.3.2.2.2 Thermoluminescent Albedo Dosimeters...................................... 29

6.3.2.2.3 Track Etch Detectors .................................................................. 30

6.3.2.2.4 Panasonic 809............................................................................ 30

6.3.2.3 Extremity Monitoring .................................................................................... 30

6.3.3 Calibration ..................................................................................................................31

6.3.3.1 Photons........................................................................................................ 31

6.3.3.2 Neutrons ...................................................................................................... 31

6.3.4 Limit of Detection........................................................................................................32

6.3.4.1 Gamma Dosimeters ..................................................................................... 32

6.3.4.1.1 Film............................................................................................. 32

6.3.4.1.2 TLD............................................................................................. 33

6.3.4.2 Neutron Dosimeters ..................................................................................... 34

6.3.4.2.1 NTA Film..................................................................................... 34

6.3.4.2.2 TLD Albedo................................................................................. 34

6.3.4.2.3 Track Etch Detectors .................................................................. 35

6.3.5 Workplace Radiation Fields ........................................................................................35

6.3.5.1 Photon Fields ............................................................................................... 35

6.3.5.1.1 Nuclear and Ramjet Engine Tests............................................... 36

6.3.5.1.2 Noble Gas Exposure................................................................... 37

6.3.5.2 Beta Particle Fields ...................................................................................... 38

6.3.5.3 Neutron Fields.............................................................................................. 39

6.3.5.3.1 Weapons-Related Neutron Fields ............................................... 39

6.3.5.3.2 Isotopic Neutron Sources............................................................ 40

6.3.5.3.3 Reactor Operations..................................................................... 40
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 5 of 121

6.3.5.3.4 Dosimetric Characteristics of Neutron Fields .............................. 40

6.4 Dose Reconstruction Recommendations ............................................................................... 41

6.4.1 Photon Radiation........................................................................................................41

6.4.1.1 Unmonitored External Radiation Photon Dose ............................................. 42

6.4.1.2 Adjustments to Recorded External Photon Radiation Dose.......................... 42

6.4.1.3 Missed Recorded External Radiation Dose .................................................. 43

6.4.1.4 IREP Input Photon Radiation Energy Fractions............................................ 43

6.4.1.5 Determination of Radiation Quantities and Exposure Geometry................... 43

6.4.1.6 Correction Factors for External Environmental Dose.................................... 45

6.4.2 Beta Radiation............................................................................................................46

6.4.2.1 Evaluation of Beta Exposure Geometries..................................................... 46

6.4.2.2 Beta-Photon Ratios for Surface Contamination ............................................ 47

6.4.2.3 Beta-Photon Ratios for Immersion in a Contaminated Cloud........................ 47

6.4.2.4 Beta-Photon Ratios for Other NTS Operational Areas.................................. 48

6.4.2.5 Estimation of Beta-Photon Exposure............................................................ 48

6.4.2.6 Skin Contamination...................................................................................... 49

6.4.3 Neutron Radiation.......................................................................................................50

6.4.3.1 Introduction .................................................................................................. 50

6.4.3.2 Neutron-to-Photon Dose Ratios ................................................................... 52

6.5 Claim Analysis Methods......................................................................................................... 53

6.5.1 Photon Dose...............................................................................................................53

6.5.2 Beta Dose...................................................................................................................55

6.5.3 Neutron Dose .............................................................................................................56

6.6 Attributions and Annotations .................................................................................................. 57

References ........................................................................................................................................64

Glossary ............................................................................................................................................74

ATTACHMENT A, DOSIMETRY TECHNOLOGY............................................................................81

ATTACHMENT B, ENERGY GROUP ALLOCATION OF PHOTON EXPOSURE 

AND PERSONAL DOSE EQUIVALENT BY NTS WORK 

AREA AND OPERATION.................................................................................103

ATTACHMENT C, BETA-PHOTON RATIO ESTIMATES FOR NTS WORK 

AREAS AND OPERATIONS............................................................................ 103

ATTACHMENT D, NEUTRON DOSES FROM ATMOSPHERIC NUCLEAR

TESTS.............................................................................................................117
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 6 of 121

LIST OF TABLES

TABLE TITLE

6-1 NTS external dosimetry–1951 to present ...............................................................................

PAGE

15

6-2 NTS film badge issue locations .............................................................................................. 19

6-3 Type of dose codes for external dosimetry............................................................................. 22

6-4 Film damage and irregularity codes ....................................................................................... 22

6-5 Characteristics of ISO reference radiation radionuclide sources used for 

neutron dosimeter calibration ................................................................................................. 32

6-6 Background dose equivalent rates ......................................................................................... 33

6-7 NTS Panasonic UD-802 detection limits for DOELAP categories........................................... 34

6-8 Primary iodines and fission product noble gases with half-lives greater 

than 1 hr................................................................................................................................. 37

6-9 Areas and operations where neutron exposure was possible................................................. 39

6-10 RBE, quality factors, or weighting factors for neutrons ........................................................... 41

6-11 NTS calculated dose to apply as co-worker dose................................................................... 42

6-12 NTS associated measured doses prior to universal badging .................................................. 44

6-13 NTS work area and operation-dependent photon fractions..................................................... 45

6-14 Nominal beta-photon ratios for operational areas at NTS....................................................... 48

6-15 Fraction of H=(10,0) by energy band for selected spectra...................................................... 51

6-16 Neutron field characteristics associated with NTS operational areas...................................... 51

6-17 Neutron-to-photon dose ratios for isotopic sources ................................................................ 52

A-1 Panasonic UD-802 dosimeter configuration ........................................................................... 86

A-2 Photon exposure to dose correction factors for NIST reference radiations ............................. 89

A-3 Bias values and uncertainties for NTS DOELAP accreditation categories .............................. 95

A-4 Fraction of HP(10,0) for neutrons with energies less than En................................................ 100

B-1 Work area and operation-dependent photon fractions.......................................................... 102

C-1 Calculated beta-photon ratios for mixed fission and activation products 

from Sedan and tests from Operations Ranger, Buster-Jangle, TumblerSnapper, and Upshot-Knothole—contaminated surfaces ..................................................... 108

C-2 Calculated beta-photon ratios for mixed fission and activation products 

from Operations Ranger, Buster-Jangle, Tumbler-Snapper, and UpshotKnothole tests—immersion in a contaminated semi-infinite cloud ........................................ 109

C-3 Photon dose constants, beta dose rates, and beta-photon ratios for NTS 

radionuclides........................................................................................................................ 110

C-4 Geometry-dependent beta-photon ratios for NTS radionuclides (Table 2-2)......................... 114
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 7 of 121

LIST OF FIGURES

FIGURE TITLE

6-1 X-ray energy dependence of PICs .........................................................................................

PAGE

21

6-2 Security badge for NRDS TLD ............................................................................................... 26

6-3 REECo badge holder ............................................................................................................. 27

6-4 REECo security badge holder with Panasonic UD-802 dosimeter.......................................... 27

6-5 Panasonic UD-807 extremity dosimeter ................................................................................. 31

6-6 Ratio of fluence to dose conversion coefficients for personal dose 

equivalent HP(10, 0°) to conversion coefficients for dose equivalent H................................... 50

A-1 Typical photon energy response curves for Li2B4O7 and CaSO4 TLD 

elements ................................................................................................................................87

A-2 Panasonic UD-802 energy responses.................................................................................... 88

A-3 NTA film neutron energy response......................................................................................... 90

A-4 Neutron energy response of NTS albedo dosimeters ............................................................. 91

A-5 Neutron energy response of NTS TED................................................................................... 93

A-6 Electrochemically etched CR-39 dose response .................................................................... 94

A-7 Lethargy neutron spectra for 252Cf calibration source with and without room 

scatter ....................................................................................................................................98

A-8 Lethargy neutron spectra for 252Cf with various thickness moderators.................................... 99

A-9 Lethargy neutron spectra for an 241AmBe calibration source with and 

without room scatter...............................................................................................................99

A-10 Fraction of HP(10,0) due to neutrons with energies less than En ......................................... 100

C-1 Beta-to-gamma dose ratios for contaminated surfaces used at Pacific tests ........................ 105

C-2 Calculated beta-photon ratios for skin for 11 atmospheric tests from 

Operation Upshot-Knothole.................................................................................................. 106

C-3 Calculated beta-photon ratios for skin for nuclear rockets and ramjet test 

effluents ............................................................................................................................... 106

C-4 Calculated beta-photon ratios for immersion in a contaminated semiinfinite cloud—maximum of values from 11 atmospheric test radionuclide 

inventories ........................................................................................................................... 110

D-1 Neutron spectra for (a) fission weapon and (b) thermonuclear weapon per 

kiloton total energy yield....................................................................................................... 118

D-2 Initial neutron dose per kT as a function of slant range from (a) fission and 

(b) thermonuclear air bursts, based on 0.9 normal sea-level air density............................... 118

D-3 Comparison of the initial neutron dose per kiloton as a function of slant 

range from fission and thermonuclear air bursts................................................................... 119

D-4 Neutron dose equivalent as a function of slant range for atmospheric 

detonations with a range of weapon yields........................................................................... 119

D-5 1-mrem neutron isodose curves for continental detonations................................................. 121
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 8 of 121

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

ALARA as low as reasonably achievable

BREN Bare Reactor Experiment Nevada

C coulomb

cm centimeter

CR-39 Columbia Resin-39 

d day

DOD U.S. Department of Defense

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

DOL U.S. Department of Labor

ECF element correction factor

EEOICPA Energy Employees Occupational Illness Compensation Program Act

EG&G Edgerton, Germeshausen and Grier

eV electron volt

GSD geometric standard deviation

H&N Holmes & Narver

HP(d) personal dose equivalent (d = depth in millimeters)

hr hour

HRA High Radiation Area

Hz hertz

IARC International Agency for Research on Cancer

ICRP International Commission for Radiological Protection

ICRU International Commission on Radiological Units and Measurements

IDO Idaho Operations Office

in. inch

ISO International Organization for Standardization

IREP Interactive RadioEpidemiological Program

keV kilovolt-electron

kg kilogram

km kilometer

kT kiloton

kV kilovolt

LANL Los Alamos National Laboratory 

LASL Los Alamos Scientific Laboratory

LLD lower limit of detection

LLNL Lawrence Livermore National Laboratory

LRL Lawrence Radiation Laboratory

m meter

MBq megabecquerel

MDL minimum detection level
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 9 of 121

MeV megavolt-electron

mg milligram

mGy milligray

mi mile

min minute

mm millimeter

MMD maximum missed dose

mR milliroentgen

mrem millirem

mSv millisievert

MT megaton

NBS National Bureau of Standards

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

NIST National Institute of Standards and Technology

nm nanometer

NPG Nevada Proving Ground

NRDS Nuclear Rocket Development Station

NTA Neutron Track Emulsion, type A

NTS Nevada Test Site

OW open window

Pan Am Pan American Airways

PIC pocket ionization chamber

POC probability of causation

QA quality assurance

QC quality control

R roentgen

RADEX controlled radiation exclusion (area)

RBE relative biological effectiveness

RCF run calibration factor

REECo Reynolds Electrical and Engineering Company

RSN Raytheon Services Nevada

SNL Sandia National Laboratories

Sv sievert

TBD technical basis document

TED track etch detector

TLD thermoluminescent dosimeter

TRU transuranic

TTR Tonopah Test Range

U.S.C. United States Code

wk week

yd yard

YMP Yucca Mountain Project
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 10 of 121

yr year

Z atomic number

μm micrometer

°C degree Celsius

°F degree Fahrenheit

§ Section
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 11 of 121

6.1 INTRODUCTION

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historic background information and guidance to assist in the preparation of 

dose reconstructions for particular sites or categories of sites. They will be revised in the event 

additional relevant information is obtained about the affected site(s). These documents may be used 

to assist NIOSH staff in the completion of the individual work required for each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of D uty.” That provision [42 U.S.C. 

§ 7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in 

the performance of duty for purposes of the compensation program if, and only if, the cancer ... was 

at least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 

(42 U.S.C. 7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. 

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

[i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally derived radiation 

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external dosimetry monitoring results are considered valid for use in dose reconstruction. No efforts 

are made to determine the eligibility of any fraction of total measured exposure for inclusion in dose 

reconstruction. NIOSH, however, does not consider the following exposures to be occupationally 

derived:

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation 

guidelines (nor the dose reconstruction regulation) define “performance of duty” for DOE employees 

with a covered cancer or restrict the “duty” to nuclear weapons work.

• Radiation from naturally occurring radon present in conventional structures

• Radiation from diagnostic X-rays received in the treatment of work-related injuries

 

1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 12 of 121

6.1.1

The purpose of this document is to describe Nevada Test Site (NTS) external dosimetry systems and 

practices. This information will be used as needed to evaluate external occupational doses for 

EEOICPA claimants. 

Purpose

Attributions and annotations, indicated by bracketed callouts and used to identify the source, 

justification, or clarification of the associated information, are presented in Section 6.6. 

6.1.2

NTS operations played an important role in the U.S. nuclear weapons program. During the period 

when the U.S. tested nuclear weapons, radiation exposure monitoring of energy employee civilian and 

U.S. military personnel associated with that testing was performed using portable radiation

instrumentation and personnel film dosimeters (NRC 1989). Operations at NTS involved atmospheric 

and underground weapons tests, experimental reactor tests intended for aircraft and rocket 

propulsion, and low-level transuranic (TRU) waste disposal (REECo 1995a; Allen and Schoengold 

1995). NTS workers received exposures from a range of fission and activation products from test 

programs conducted beginning in the early 1950s. Exposure venues of particular significance include 

sample and measurement device recovery to assess weapon yields, terrestrial and airborne fallout 

plume tracking, aircraft operations (when used) for weapon assembly deployment and surveillance, 

and postevent decontamination processes [1]. 

Scope

Radiation monitoring and control programs instituted with the mission of NTS included personal 

dosimetry, area monitoring, source term characterization, and measurements of fallout 

(contamination) dispersion [1]. As NTS test programs progressed, efforts to measure exposures and 

limit dose improved (Allen and Schoengold 1995; DeMarre 2002). The atmospheric nuclear test 

series consisted of a number of operations between 1951 and 1958, and again from 1961 to 1963

(DOE 2000). Each operation consisted of a number of individual tests. Underground nuclear testing 

occurred at NTS as early as 1958 and continued to 1992 (DOE 2000). 

This TBD contains supporting documentation to assist in the evaluation of occupational external 

doses from these processes using the methodology in the External Dose Reconstruction

Implementation Guidelines (NIOSH 2006). NIOSH considers the available data and methods for 

performing external dose reconstruction to be adequate for estimating with sufficient accuracy the 

external radiation doses at NTS from 1951 to the present. 

6.2 DOSIMETRIC BASIS OF COMPARISON

Since the start of the Manhattan Engineer District program in the early 1940s, various dosimetric 

concepts and quantities have been used to measure and record occupational exposure from external 

radiation sources. The selection of the measurement quantities to be used for radiation protection 

was initially based on the radiation interaction properties of the primary radiations of interest – 

photons (X- and gamma rays), electrons (beta particles), and neutrons – and the measurement

methods employed. Sections 6.2.1 to 6.2.3 discuss these quantities briefly.

The problem with the use of different quantities for different radiations is that it prevented direct 

comparison of measurements of the three primary radiation qualities. In the 1950s the industry 

recognized a need for a special quantity to facilitate comparison of measurements of the various 

radiation qualities. The concepts of quality factor and dose equivalent were formally introduced by the 

International Commission for Radiological Protection (ICRP) and the International Commission on 

Radiation Units and Measurements (ICRU) in 1962 (ICRU 1962). The special unit of dose equivalent
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 13 of 121

was the rem. As early as 1961, NTS radiation exposure criteria and guidance were given in rem 

(AEC 1961).

In 1985, the ICRU defined a new set of operational quantities defined as radiation quantities for 

operational radiation protection measurement purposes (ICRU 1993). These quantities have a 

common definition for the three primary radiation qualities, so they have the advantage that they 

provide a means of direct comparison of measurements for these radiations.

The operational quantity recommended for individual or personal monitoring is the personal dose 

equivalent HP(d), where d is the depth (in millimeters) and represents the point of reference for dose 

in tissue. For weakly penetrating radiation of significance to skin dose, d is 0.07 mm and the 

operational quantity is noted as HP(0.07). For strongly penetrating radiation of significance to wholebody dose, d is 10 mm and the operational quantity is noted as HP(10). Both HP(0.07) and HP(10) are 

recommended as the operational quantities to be recorded for radiological protection proposed by the 

ICRU (1993).

These personal dose equivalents, HP(0.07) and HP(10), have been used in the DOE Laboratory 

Accreditation Program (DOELAP) for accreditation of the Department’s personnel dosimetry systems 

since the 1980s (DOE 1986a). The International Agency for Research on Cancer (IARC) Three 

Country Combined Study (Fix et al. 1997; Fix, Wilson, and Baumgartner 1997) and IARC 

Collaborative Study (Thierry-Chef et al. 2002) selected HP(10) as the quantity to assess error in 

recorded whole-body dose for workers in IARC nuclear worker epidemiologic studies.

6.2.1

From the beginning of operations at NTS, exposure was used as the basis of photon measurement. 

When in italics, the term exposure designates a radiation measurement quantity based on the 

electrical charges created in air due to interaction of photons. It is specific to air and photon

measurement. The unit of exposure is coulomb per kilogram, and the special unit of exposure is the 

roentgen (1 R = 2.58 × 10-4 C/kg). Although the term exposure dose was used in the mid-1950s, 

exposure is not dose because it is not a measure of energy deposition in mass of material. The ICRU 

recommended the use of exposure in 1962 (ICRU 1962). However, dose in soft tissue can be 

determined from exposure by the following relationship,

Photon Measurement Quantities

D = fX

where f is a constant that relates the exposure X in roentgen for a given photon energy to dose in soft 

tissue, and was typically taken to be 0.877 rad/R. Because 1 R is numerically slightly greater than 

1 rad, it is favorable to claimants to assume a numerical equivalence of the two quantities. Therefore, 

if the energy and hence the constant f of the exposing photon radiation are unknown, it is appropriate 

to assume that 1 R equals 1 rad, which equals 10 mGy. 

Instrument and dosimeter calibrations, dose measurements, and dose records were made in terms of 

exposure in units of milliroentgen. However, as noted above, from 1961 quarterly and annual limits 

were specified in millirem (AEC 1961), where the general term dose was used without specific 

reference to dose equivalent. As a result, individual monitoring results at NTS were recorded in 

millirem, although the measurements had been made in terms of exposure. However, for photons, 

the values of exposure and dose equivalent were considered to be essentially the same (Griffith 2004; 

Brady and Iverson 1968). In effect, a de facto conversion factor of 1 rem/R was used for dose 

recording purposes.

Until July 1970, individual monitoring at the Nuclear Rocket Development Station (NRDS) was 

provided by the Reynolds Electrical and Engineering Company (REECo) using the standard NTS film 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 14 of 121

dosimeter (Boone, Bennett, and Adams 1970). From July 1970 until NRDS operations ceased in 

January 1973, individual monitoring was conducted by Pan American World Airways (Pan Am) using 

thermoluminescent dosimeters (TLDs) (Boone, Bennett, and Adams, 1970). However, throughout 

NRDS operations, dosimeters were calibrated in terms of exposure and doses were recorded in 

millirem.

Beginning in 1987, NTS occupational exposures were recorded in terms of personal dose equivalent, 

HP(d). For exposures during the period from 1962 to 1986, dose reconstructors should use the 

recorded photon dose values in terms of exposure, together with the Exposure to Organ Dose 

coefficients in Appendix B of NIOSH (2006) to determine organ dose. Since 1987, the recorded 

values are in terms of HP(10), and the Deep-Dose-Equivalent-to-Organ-Dose conversion factors (of 

NIOSH 2006, Appendix B) should be used.

6.2.2

As was common practice, NTS beta particle measurements were made in terms of absorbed dose, D

[2]. Until introduction of the ICRU-defined operational quantities in 1985, beta doses were recorded 

as millirad. Because the quality factor for electrons (beta particles) was set at 1, the absorbed dose 

values are considered numerically equivalent to dose equivalent (ICRU 1993). 

Beta Measurement Quantities

6.2.3

The basis for comparison for neutron radiation is complex because, historically, the calibration of 

dosimeters to measure neutron dose was based on different dosimetric quantities (such as first 

collision dose and multiple collision dose). However, the neutron dose equivalent specified by the 

National Council on Radiation Protection and Measurements (NCRP) has been used since 1971. 

Evaluation of the numerical difference in comparison with the HP(10) dose used in DOELAP 

performance testing is used to establish relative values of the dose conversion factors for the dose 

quantities in conjunction with characteristics of the neutron dosimeter response characteristics and 

workplace radiation fields.

Neutron Measurement Quantities

Dose reconstructors should convert recorded neutron dose to HP(10) using the bias values that 

appear in Table 6-1, and use the Deep-Dose-Equivalent-to-Organ-Dose conversion factors from 

Appendix B of NIOSH (2006) to calculate the appropriate organ doses.

6.3 DOSE RECONSTRUCTION PARAMETERS

Overall accuracy and precision of the original recorded individual worker doses and their 

comparability to be considered in using NIOSH (2006) guidelines depend on the following factors (Fix 

et al. 1997; Fix, Wilson, and Baumgartner 1997):

• Administrative practices adopted by facilities to calculate and record personnel dose based on 

technical, administrative, and statutory compliance considerations.

• Dosimetry technology, which includes physical capabilities of the dosimetry system such as 

the response to different types and energies of radiation, in particular to mixed radiation fields.

• Calibration of monitoring systems and similarity of the methods of calibration to sources of 

exposure in the workplace.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 15 of 120

Table 6-1. NTS external dosimetry–1951 to present (DeMarre 1993; Allen and Schoengold 1995; Bechtel Nevada 2001). Photon- beta dosimeters

Dates Operation Dosimeter Description Issue and exchange

Measurement 

quantity Bias GSD MDL MMD

1/27/51–

2/6/51

Ranger, 

1/27/51–

2/6/51

DuPont 552 

Packet

DuPont 552 packet, including:

Type 502 low-range element (0.05 to 10 R)

Type 510 high-range element (5 to 50 R)

Brass/cadmium filters, 0.020-in. thick 

(symmetrical coverage on both sides with 

open area) (Shipman et al. 1951)

Issued to personnel who entered 

radiation areas and air crews. 

Exchanged daily. Supplemented 

with self-reading pocket dosimeters 

for exposure control. (Sometimes 

under-responded relative to film.)

Photon: 

Exposure

1.1 1.23a 40 mR 5.0 R

10/22/51–

11/29/51

Buster–

Jangle, 

10/22/51–

11/29/51

DuPont 553 

Packet

DuPont 553 packet, including:

Type 502 low-range element (0.02 to 10 R)

Type 510 high-range element (5 to 50 R)

Type 606 high-range element (10 to 300 R)

Brass/cadmium filters, 0.020-in. thick 

(symmetrical coverage on both sides with 

open area). Lead filters, 0.020-in. thick, 

possibly used with badges issued to Camp 

Desert Rock personnel (Kean 1951; Shipman 

et al. 1951; Storm 1951)

Issued to personnel who entered 

radiation areas, air crews, and 

people with potential for exposure 

from experiments.

Exchanged daily 

Self-reading pocket dosimeters 

sometimes used.

Badges for Camp Desert Rock 

personnel were exchanged 

following deployment

Photon: 

Exposure

1.1 1.23a 40 mR 5.0 R

4/1/52–

6/5/52

Tumbler–

Snapper, 

4/1/52–

6/5/52

DuPont 558 

Packet DuPont 558 packet, including:

Nevada Proving Ground Personnel

Lead filters, 0.028-in. thick, symmetrical on 

both sides (except first test - ABLE)

Type 508 low-range element (0.01 to 6 R)

Type 1290 high-range element (20 to 3000 R)

(Brady and Nelson 1985)

Issued to NPG test participants. 

Exchanged daily

Photon: 

Exposure

1.6 1.23a 40 mR 5.0 R

DuPont

Type 502 LASL badge

Debris cloud sampling pilots

Type 502 low-range element (0.02 to 10 R)

Brass/cadmium filters, 0.020-in. thick 

(symmetrical coverage on both sides with 

open area)

Exchanged daily Photon: 

Exposure

1.4 1.19a 40 mR 5.0 R

1953–

June 1960

All of NTS

Upshot–

Knothole, 

3/17/53–

6/4/53

DuPont 559 

Packet

DuPont 559 Packet, including:

Type 502 low-range element (0.02 to 10 R)

Type 606 high-range element (10 to 300 R)

Lead filters, 0.028-in. thick (symmetrical 

coverage on both sides with open area)

(Brady and Nelson, 1985; Collison 1953)

Exchanged daily (Exposures might 

have included prompt radiation, 

including neutrons.)

Photon: 

Exposure

0.9 1.28a 40 mR 5.0 R

All of NTS

Teapot, 

2/18/55–

5/15/55

DuPont 559 packet, including:

Type 502 low-range element (0.02 to 10 R)

Type 606 high-range element (10 to 300 R)

Lead filters, 0.028-in. thick (symmetrical 

coverage on both sides with open area)

(Collison 1955)

Issued to essentially all test 

participants.

Exchanged daily

(Exposures might have included 

prompt radiation, including 

neutrons.)

Photon: 

Exposure

0.9 1.28a 40 mR 5.0 R

All of NTS

Plumbob, 

4/24/57–

10/07/57

DuPont 559 packet, including:

Nevada Test Organization

Type 502 low-range element (0.02 to 10 R)

Type 606 high-range element (10 to 300 R)

Lead filters, 0.028-in. thick (symmetrical 

coverage on both sides with open area)

Exchanged monthly and following 

radiation area work.

Photon: 

Exposure

1.0 1.19a 40 mR 0.240 R
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 16 of 120

Photon- beta dosimeters

Dates Operation Dosimeter Description Issue and exchange

Measurement 

quantity Bias GSD MDL MMD

All of NTS,

Plumbob, 

4/24/57–

10/07/57

DuPont 559 Packet, including:

Camp Desert Rock personnel

Type 502 low-range element (0.02 to 10 R)

Type 606 high-range element (10 to 300 R)

Aluminum, copper, and tin/lead laminate 

filters with open area.

Exchanged at various intervals. Photon: 

Exposure

1.0 1.19a 40 mR 5.0 R

All of NTS

Hardtack 2, 

9/19/58–

10/30/58

DuPont 559 packet, including:

Type 502 low-range element (0.02 to 10 R)

Type 834 high-range element (5 to 800 R)

Lead filters, 0.028-in. thick (symmetrical 

coverage on both sides with open area). 

Packet and filter enclosed in 0.004-in.-thick 

plastic bag.

Exchanged monthly and on exit 

from radiation areas if ≥100 mR 

was suspected.

Photon: 

Exposure

1.0 1.19a 40 mR 5.0 R

July 1960 

–1965 

All of NTS

Dominic II 

(Sun-beam), 

7/7/62–

7/17/62

DuPont 301-

4 Packet 

(Also known 

as DuPont 

Type 556)

DuPont 301-4 packet, including:

Type 508 low-range element (0.03 to 5 R)

Type 834 high-range element (5 to 800 R)

Lead filters, 0.028-in. thick (symmetrical 

coverage on both sides)

Packet covered with 0.004-in.-thick plastic 

bag.

Exchanged monthly for general 

exposures and on exit from 

radiation areas for exposures likely 

to exceed 100 mR.

Photon: 

Exposure

1.0 1.23a Photon: 40 

mR

Photon:

240 mR

1966–Feb. 

1971

All of NTS DuPont 

Type 556 

Packet

DuPont Type 556 film pack 

Type 508 (519 also referenced) low-range 

element (0.03 to 5 R)

Type 834 high-range element (10 to 1,000 R)

Four-area filter described: tantalum-cadmium, 

tantalum, Teflon, open

With fast neutron pack, dosimeter was 

sensitive to mixed fields with thermal and fast 

neutrons, X-rays, beta, and gamma

Exchanged monthly for general 

exposures and on exit from 

radiation areas for exposures likely 

to exceed 100 mR.

Photon: 

Exposure

Beta:

Absorbed 

dosec

1.0 1.23a,b Photon: 40 

mR

Beta: 

40 mremd

Photon:

240 mR

Beta:

240 mrem

March 

1971–

1986 

All of NTS Kodak Type 

III

Low-range element (0.03 to 10 R)

High-range element (10 to 800 R)

Exchanged monthly for general 

exposures and on exit from 

radiation areas for exposures likely 

to exceed 100 mR.

Photon: 

Exposure

Beta:

Absorbed 

dosec

1.0 1.23b Photon:

30 mRe 

Beta: 

30 mreme

Photon:

180 mR

Beta:

180 mrem

1970–

1972

NRDS Pan Am 

TLD

Two-element LiF, Type 700 Quarterly Photon: 

Exposure

Beta:

Absorbed 

dosec

0.8 1.23b Photon:

15 mRf

Beta: 

15 mremf

Photon:

30 mR

Beta:

30 mrem

1987–

Present

All of NTS Panasonic 

802

Four-element TLD – two Li2B4O7:Cu chips 

and two CaSO4:Tm chips. Filtration provided 

to determine gamma, deep dose, shallow 

dose, beta, and eye dose. NTS badge holder 

used from 1987. Used for beta gamma only.

Quarterly Personal dose 

equivalent, 

HP(d)g

0.9 1.23b Photon:

30 mrem

Beta:

25 mrem

Photon:

60 mrem

Beta:

50 mrem
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 17 of 120

Photon- beta dosimeters

Dates Operation Dosimeter Description Issue and exchange

Measurement 

quantity Bias GSD MDL MMD

2001–

present

All of NTS Panasonic 

809

combination 

dosimeter

Multielement TLD containing four elements.

E1 - gamma-sensitive 7Li2

11B4O7(Cu), 

enriched to 99.99% in 7

Li.

E2, E3 and E4 – neutron-sensitive 6

Li2

10B4O7(Cu) chips. Li-6 enriched to 95.33% 

and B-10 enriched to 94.64%. Elements are 

shielded with tin and cadmium on front and 

back, in various combinations. Issued for 

mixed beta, gamma, neutron.

Quarterly Personal dose 

equivalent, 

HP(d)g 

0.9 1.23b Photon:

30 mremf 

Beta:

25 mremf 

Photon:

60 mrem

Beta:

50 mrem

Neutron dosimeters

1961–

1979

Areas and 

operations 

where the 

potential for 

neutron 

exposure 

existed

Kodak NTA Responds to neutrons with energies above 

0.8 MeV; range under near-ideal conditions 

0.1 to few rem of neutrons; high gamma 

doses might mask neutron tracks

Exchanged monthly for general 

exposures and on exit from 

radiation areas for exposures likely 

to exceed 100 mR.

Dose 

equivalenth

0.5.i 1.52b 100 

mremi,j

600 mrem

1979–

1986

Albedo 

Dosimeter

Hankins-type albedo dosimeter. Consists of 

four pairs of TLD-600 and TLD-700 (6LiF and 7

LiF) in cadmium pillbox for thermal neutron 

suppression. High sensitivity to low-energy 

neutrons, with decreasing response as 

energy increases.

Monthly

Issued only to individuals with 

potential for exposure to neutrons

Dose 

equivalenth

1.0.

i 1.23b 20 mremi,j 120 mrem

1987–

2000

TED Three pieces of CR-39 plastic used to detect 

neutrons with energies above 100 keV. 

Quarterly, except for limited 

number of workers (radiographers, 

well loggers, and personnel 

routinely entering HRAs).

Issued only to individuals with 

potential for exposure to neutrons

Dose 

equivalenth

0.9i 1.23b 50 mremi,j 100 mrem

2001–

present

Panasonic 

809 

combination 

dosimeter.

Multielement TLD containing four elements. 

E1 - gamma-sensitive 7

Li2

11B4O7(Cu), 

enriched to 99.99% in 7

Li.

E2, E3 and E4 – neutron-sensitive 6

Li2

10B4O7(Cu) chips. Li-6 enriched to 95.33% 

and B-10 enriched to 94.64%. Elements are 

shielded with tin and cadmium on front and 

back, in various combinations. Includes CR39.

Quarterly, except for limited 

number of workers (radiographers, 

well loggers, and personnel 

routinely entering HRAs).

Issued only to individuals with 

potential for exposure to neutrons

Dose 

equivalenth

1.0.

i 1.23b 40 mremi,j 80 mrem 

a. Based on uncertainty values provided by NRC (1989) and ORAUT (2006a, Equation 4-1) for the 95th-percentile estimate.

b. Based on anassumption favorable to claimants of a sigma (standard deviation) equal to ±20% and 95th-percentile values enveloped by 2 sigma.

c. Numerically equivalent to dose equivalent (Q = 1).

d. Assumed to be the same as that for DuPont Type 502 and 508 films (NRC 1989).

e. From DeMarre (2002). 

f. Assumed to be the same as Panasonic 802 dosimeter.

g. Deep dose equivalent = HP(10).

h. Source: NCRP (1971).

i. See discussion in Section 6.4.3.1.

j. Based on reported values, corrected for potential energy-dependent under-response (NTA film and Panasonic TED) and ratio of the conversion coefficients for personal dose equivalent, 

HP(10), to those for dose equivalent, H, (NCRP 1971) in the 100–2,000 keV energy range. 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 18 of 121

• Workplace radiation fields that could include mixed types of radiation, variations in exposure 

geometries, and environmental conditions.

Examination of the beta and photon (X- and gamma rays) radiation type, energy and geometry of 

exposure in the workplace, and characteristics of the dosimeter response is crucial to assessment of 

bias and uncertainty of the original recorded dose. Table 6-1 lists the chronology of NTS external 

dosimetry methods and practices beginning in 1951. The parameters of significance are the bias, 

uncertainty, minimum detectable level (MDL), and potential maximum missed dose (MMD). For the 

purpose of this document, Bias is defined as the ratio of the true dose to the reported dose (i.e., if the 

reported dose was an overestimate of the true dose, Bias <1). The values of uncertainty are 

presented as the geometric standard deviation (GSD). The values of Bias and GSD for the period 

from 1951 to 1966 are based on the values of bias and uncertainty reported in Film Badge Dosimetry 

in Atmospheric Nuclear Tests (NRC 1989) for the dosimetry systems and practices in use during the 

period of atmospheric testing.

The bias factors listed in Table 6-1, are the factors by which recorded values in the dosimetry records 

should be divided to provide the best estimate of the measurement quantity [3]. The value of the 

measurement quantity should then be multiplied by the dose conversion factors in Appendix B of 

NIOSH (2006) to obtain the organ dose [4]. The bias values for neutrons include an estimate of 

potential under-response due to the inherent energy response of the detector and spectral differences 

between the calibration sources and operational spectra. Potential detector over-response (e.g., for 

TLD albedos) maintains favorability for claimants.

The uncertainty in the best estimate of the measurement quantity is accounted for in the GSD. The 

GSD is discussed in more detail in ORAUT (2006a) and defined in Equation 4-1 of that document 

presented as follows:

)()( 1.65585

1

50th percentile

95th percentile GSD = (6-1)

Bias and uncertainty for current DOE dosimetry systems are well documented for HP(0.07) and HP(10)

under DOELAP (DOE 1986a,b, 1995a). The performance of current dosimeters can be compared to 

performance characteristics of historical dosimetry systems in the same, or highly similar, workplaces. 

In addition, current performance testing techniques can be applied to earlier dosimetry systems to 

achieve a consistent evaluation of those systems. Dosimeter response characteristics for radiation 

types and energies in the workplace are crucial to the overall analysis of error in recorded dose.

The MDL is typically established at the point where the laboratory uncertainty of the readings at the 

95% confidence level is ±100% in normal distribution terms. The MMD is equal to one-half the MDL 

multiplied by the number of exchange or monitoring periods (NIOSH 2006). With an MDL = 0.04 R, 

and 250 monitoring periods (5 days × 50 weeks), the MMD = 0.04/2 × 250 = 5 R.

6.3.1

When the testing program at NTS began in January 1951, the Los Alamos Scientific Laboratory [now 

named Los Alamos National Laboratory (LANL)] was responsible for administering the external 

dosimetry program (Allen and Schoengold 1995). While contractor organizations and the military 

were involved in issuing and collecting badges for some of the early operations at NTS, LANL 

performed calibration, processing, and interpretation work (Allen and Schoengold 1995). In July 

Administrative Practices
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 19 of 121

1955, REECo assumed responsibility for most onsite radiological safety functions (Allen and 

Schoengold 1995). 

Until 1966, there was no determination of shallow or skin dose from the film badges (DeMarre 2002; 

Allen and Schoengold 1995). With the introduction of a new multi-element film dosimeter in 1966 until 

the conversion to TLDs in January 1987, the open window (OW) was used to make a separate 

determination of shallow dose and deep dose (Brady and Iverson 1968; DeMarre 2002; Allen and 

Schoengold 1995). Shallow dose was determined by comparing the OW and closed-window readings 

and, therefore, did not include the penetrating photon component. This same type of determination 

was used for TLDs used by Pan Am at the NRDS from 1970 to 1972 (Pan Am 1967; Boone, Bennett 

and Adams 1970). On the NV-185 form used for dose records at NTS, the shallow dose was 

incorrectly called the "Skin of the Whole-Body" dose when it was, in fact, only the beta and low-energy 

photon components. Beginning in 1987, procedures used currently for calculating shallow or skin 

dose were adopted (Allen and Schoengold 1995). 

Beginning in 1987, with the introduction of Panasonic TLDs, the deep (or whole-body) dose and the 

shallow (or skin) dose were reported, but the shallow dose included the penetrating photon 

component and can therefore be considered the total skin dose (DeMarre 2002; Allen and 

Schoengold 1995). 

Prior to 1987, unexposed control films were processed together with the personnel dosimeters [5]. 

Two sets of control films were used. Personnel films were processed together with two unexposed 

films from the same emulsion series that had been stored at the Dosimetry Laboratory in Mercury. 

Additional unexposed “area” films that had been stored at the NTS issue locations (Table 6-2) were 

processed as well. Additional badges could be obtained at Building 1000.

Table 6-2. NTS film badge issue locations (REECo 1961).

Area Building

Mercury 111, 155

Yucca and Frenchman’s Flats CP-2

12 Rad-safe trailers and stations

400 Rad-safe stations

401 Rad-safe stations

Readings from the Mercury control films were subtracted from the dosimeter readings to obtain a net 

reading for determining exposure (REECo 1961, 1962). The net reading represented the 

occupational exposure plus the differential between the Mercury and work area environmental levels. 

If there was an indication that there had been a problem with use or storage of the films at the issue 

locations, the “area” controls were used for background subtraction. However, such incidents were 

rare, and were noted in the dosimetry records when they occurred.

It is important to note that, before 1986, the film badge was used for the dose of record. If the film 

badge was lost or damaged, the health physicists preferred to use cohort dosimetry results and 

ambient radiation levels. Pocket dosimeter results could be a consideration, but they were only used 

for short work periods and might not have been dependable because they were potentially subject to 

discharge (DeMarre 2006a).

Beginning in 1987, with the introduction of TLDs, the procedure was continued with TLD background 

dosimeters (REECo 1990; Allen and Schoengold 1995). 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 20 of 121

6.3.1.1 Exposure Limits

From 1951 to 1958, the allowable external exposure limits for occupational workers at NTS were 

generally consistent with NCRP recommendations (Allen and Schoengold 1995). During the 1951 to 

1952 test series, participants could receive up to 3 R of gamma exposure for a 13-wk period (Allen 

and Schoengold 1995). Pilots and crew could receive up to 3.9 R of exposure (Allen and Schoengold 

1995). For the 1953 and 1955 series, workers could receive up to 3.9 R (Allen and Schoengold 

1995). Beginning in 1957, the maximum permissible exposure for test participants was limited to 3 

rem per 13-wk period and 5 rem per calendar year (Allen and Schoengold 1995). In 1961, the U.S. 

Atomic Energy Commission (AEC) Standard Operating Procedure, Nevada Test Site Organization 

(NTSO), Chapter 0524, Radiological Safety (AEC 1961), stated that the radiation exposure criteria for 

NTS personnel were 3 rem per quarter and 5 rem/yr. However, with the approval of the Test 

Manager, an NTS worker could receive as much as 12 rem/yr. DOE (1982), Radiological Safety, also 

set 3 rem per quarter and 5 rem/yr as limits for exposure for occupational workers.

6.3.1.2 Dosimeter Exchange

The dosimeter exchange procedures evolved over the period from 1951 to 1987. From 1951 through 

1954, 1-d film dosimeters were issued to personnel entering a controlled radiation exclusion (RADEX) 

area. The badges were collected when leaving that area. Entry to a RADEX area was not allowed 

without a film dosimeter (Shipman 1953). In 1955, there was a transition to a combination of 1-d, 1-

wk dosimeters for working in RADEX areas. In 1956, 1-wk dosimeters were issued for exclusion 

areas and by December, monthly dosimeters were in use. In 1957, the regular practice was to issue 

monthly dosimeters. Prior to 1957, the majority of workers were not badged because their job 

responsibilities did not require that they enter RADEX areas. 

On April 1, 1957, the issue card system came on line (DeMarre 2002). All persons entering the NTS 

had to have a film badge, with the correct monthly color coding on the exterior of the badge. An 

individual working in a RADEX area could have many more than 12 badges for the year. If the pocket 

dosimeter indicated a possible exposure on the work shift, the film dosimeter was pulled and 

replaced. From April 1957 through the end of testing, in 1992, all NTS personnel had dosimeters 

(DeMarre 2002). The front gate security officer would check to make sure each badge had the proper 

color. If not, it was necessary to go to the Badge Office in Building 1000 at the gate to get a current 

badge.

Until 1987, the film badges were exchanged monthly for all individuals, or on exit from radiation areas 

if an exposure of 100 mR or more was measured (or suspected) (Allen and Schoengold 1995). In 

addition to film badges, self-reading pocket dosimeters were issued to persons entering RADEX 

areas, which were controlled locations at which an exposure was usually expected. The purpose of 

issuing pocket ionization chambers (PICs) to persons entering a RADEX area was to provide an 

action alert. The maximum reading on the pocket dosimeters most commonly used was 200 mrem. 

The pocket dosimeter could discharge when dropped, knocked, or exposed to water. Unless there 

was good evidence that the pocket dosimeter was damaged, a high reading of a PIC ( ≥ 100 mrem) 

triggered the action of collecting and processing the personnel dosimeter being used at the time 

(DeMarre 2006a). PIC results were not used unless the personnel dosimeter had somehow been 

compromised. The PIC result would have been included in a special investigation of the incident. 

PIC results were only used as dose-of-record when film badge results were not available, because of 

their typical over-response characteristics (Figure 6-1).

Following the introduction of TLDs in January 1987, dosimeters were issued on a quarterly basis 

unless a particular job assignment indicated the need for more frequent issue and readout (REECo 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 21 of 121

Figure 6-1. X-ray energy dependence of PICs (Sanders, Auxier, and Cheka 1959).

1990; Allen and Schoengold 1995). Measured exposures were added to the yearly and quarterly 

accumulated exposures (REECo 1990; Allen and Schoengold 1995). 

6.3.1.3 Dosimetry Codes for External Monitoring

Employer codes and job titles for NTS contractors [REECo, Edgerton, Germeshausen and Grier 

(EG&G), Holmes & Narver (H&N), and Raytheon Services Nevada (RSN)] are available to dose 

reconstructors on the O: Drive maintained by the NIOSH Office of Compensation Analysis and 

Support.

Current NTS contractors include Bechtel Nevada; Lockheed Martin Nevada Technologies, Inc.; 

Johnson Controls Nevada, Inc.; and Wackenhut Services, Inc [6]. Former NTS contractors included 

H&N (1956–1990), Fenix & Scisson of Nevada (1963–1990), EG&G (1951–1995), REECo (1953–

1995), 

and RSN (1990–1995) [6]. EG&G had offices and shops in Las Vegas, but some EG&G workers 

worked at NTS for extended periods. Some REECo workers were assigned to the Tonopah Test 

Range (TTR); Sandia National Laboratory is the custodian of TTR dosimetry records.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 22 of 121

The computerized external dosimetry records contain the codes listed in Tables 6-3 and 6-4 (DeMarre 

2003). Body part codes can be found in Table A-2 of ORAUT (2006b). However, the important body 

part codes for external dosimetry are 01 WB Whole body; 02 SK Skin; 03 LH L-HAND; and 04 

RH R-HAND.

Table 6-3. Type of dose codes for external 

dosimetry. 

Code Definition

00000001 Dose type unknown

00000002 Gamma 

00000003 Beta

00000004 Thermal neutron 

00000005 Other neutron (other than thermal)

00000006 Body part-1

00000007 Body part-2

00000008 Initial gammaa

00000009 Initial neutrona

a. Refers to initial radiation at the time of 

detonation, and not normally used for workers 

since they were not present at the time.

Table 6-4. Film damage and irregularity codes. 

Damage code Description

A Lost dosimeter

B Light damage

C Heat damage

D Pressure damage

E Factory damage

F Processing damage

G Medical exposure

H Non-personnel exposure X-ray

I Destroyed

J Water damage

K Age damage

L Undetermined damage

N Occupation damage

O Non-returned

X Dose by investigation

Y Late return

Additional information regarding recordkeeping practices during the period from 1945 to 1962 can be 

found in DeMarre (2006b).

6.3.1.4 Duplicate Dosimetry Records

Dosimetry for workers at NTS presents a particular problem for dose reconstruction because of the 

large number of DOE facilities where workers were involved with NTS activities. The standard 

operating practice at NTS called for wearing only the NTS personnel dosimeter (film or TLD) while on 

site (REECo 1995a). This posed no problem for NTS contractor employees assigned to the site who 

were routinely issued personnel dosimeters that served to provide the dose of record. Visitors, 

contractor 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 23 of 121

employees from other sites, and temporary workers were also issued NTS dosimeters if called upon 

to work on site (REECo 1995a). However, major laboratories such as LANL, Lawrence Livermore 

National Laboratory (LLNL), Sandia National Laboratories (SNL), and their contractors (e.g., R E 

McKee) were heavily involved in NTS activities. Such temporary assignments might be of several 

weeks duration, perhaps punctuated by a return home over a weekend or for a few days [7]. 

Normal practice called for temporary or casual workers from other AEC/DOE contractor sites to obtain 

their NTS personnel dosimeter/security badge at the badge house in exchange for their regular, 

employer-issued dosimeter and security credential. Thus, a worker on temporary assignment from, 

for example, LLNL would come to the main gate where they would exchange their Livermore-issued 

personnel dosimeter/security credentials for their NTS-issued equivalents. When the workers exited 

the site at the end of their temporary assignments, they would turn in their NTS-issued dosimeters in 

exchange for the dosimeters they had left at the badge house (REECo 1995a), [8]. 

This system worked well, but was not foolproof and, thus, there exists a small but real possibility that 

in some cases the NTS dosimeter and

As a result, some individuals would have dosimetry records maintained both at their parent 

laboratory’s dosimetry service as well as those operated by NTS. In fact, there are cases in which 

duplicate records exist for the same exposure. This was particularly true in the period prior to the time 

that REECo assumed responsibility for NTS activities (July 1955). For example, there have been 

LANL-based employees who reported to NTS for the Ranger Operation or the Buster-Jangle 

Operation in 1951. LANL performed the dosimetry for NTS for those two operations and there are 

copies of the same data in both the LANL and NTS record systems [9]. 

a dosimeter from the worker’s parent facility were both worn. 

Such might be the case if workers were flown into or out of the alternative air strip at Yucca Flats or in 

Area 17, and hence did not make the badge exchange at the Main Gate. In some cases, workers 

might have been issued a special dosimeter by the parent dosimetry laboratory as a means of 

tracking a special exposure away from the home facility. This was done from time to time when the 

potential for a high or unusual exposure was thought to be present, and the parent facilities desired a 

separate measurement that, if necessary, could be processed immediately upon return from NTS. 

This would provide an immediate indication of the worker’s exposure status. There were also 

occasions when Lawrence Radiation Laboratory (LRL)/LLNL personnel, for example, were provided 

with test dosimeters for various reasons (REECo 1995a). 

There were a few instances in which personnel wore duplicate dosimeters [9]. Although care was 

taken to avoid duplicate dose assignment, the obvious problem associated with double-badging is that 

the individual could be assigned the same dose twice. While this is favorable to claimants, it could 

result in a situation in which the unearned dose assigned to an individual might be sufficient to result 

in an unduly and inappropriately high POC leading to unwarranted compensation.

In situations where an employee of one of the laboratories obtained an exposure that appears in the 

NTS records system, the dose reconstructor should review laboratory records for the same period to 

determine if the same exposure for the same period appears in those records as well. If so, a note 

should be made that the exposure was received at NTS, and the exposure appearing in laboratory 

records should not be included.

Double-badging and associated doubled dose assignments are likely to be difficult to establish with 

any degree of certainty. Unless there is unequivocal documentation that two badges were issued and 

that two doses were assigned for the same period, the prudent action would be for the dose evaluator 

to assign all recorded dose to the claimant. However, if two personnel dosimeters were assigned to 

the same person and both were worn over the same or approximately the same period with 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 24 of 121

approximately the same recorded dose, this is prima facie evidence that the person was in fact 

double-badged and highly suggestive that a doubled dose assignment was made. In such instances, 

a determination should be made whether this was likely the case. If so, only the higher of the doses 

from the two dosimeters should be assigned to the claimant.

Similarly, if documentation exists that a person was in fact double-badged, with the time frame for one 

badge overlapping the time frame for the other badge at both ends, the dose from the former 

dosimeter should be assigned unless it is the smaller of the two. This is illustrated by the following 

hypothetical example: Worker ABC was issued a personnel dosimeter on January 15, 19xx, by his 

parent facility and another by NTS on January 20, 19xx. The NTS dosimeter was turned in on 

January 30, 19xx, and from it a dose of 400 mrem was assigned. The dosimeter from the parent 

facility was turned in on January 31, 19xx, and indicated a dose of 600 mrem. Documentation 

indicates that both badges were worn or carried on the person while at NTS. In this case, the 

claimant should be assigned a dose of 600 mrem for the period from January 15 through January 31. 

If it cannot be established that both dosimeters were worn or carried while at NTS, the readings from 

both dosimeters should be assigned to the claimant.

An additional concern arises because there were some REECo employees who were hired and 

terminated at NTS, but who were assigned to offsite locations. For example, REECo employees were 

assigned to support SNL at TTR, a situation that was possible until the mid-1990s [10]. Dose 

reconstructors should be aware that the only records NTS would have are the dosimetry results for 

the in- and out-processing period (usually 1 d each). If mention is made of working at TTR, the dose 

reconstructor might need to request additional information from SNL (e.g., either the dosimetry data or 

a clear statement that the employee was not monitored). If the TTR employment was not verified by 

DOL and no dosimetry records for TTR are included in the file, the TTR employment should not be 

included in the dose reconstruction and the entire period of DOL-verified employment should be 

evaluated as NTS employment.

6.3.1.5 Special Issues

6.3.1.5.1 External Exposure to “Hot Particles”

Highly radioactive particles are produced by some of the operations at NTS (e.g. atmospheric testing 

or reactor operations such as the nuclear rocket tests) (NRDL 1968; NCRP 1990). The size of hot 

particles contained in nuclear fallout ranges from 10 nm to 20 μm for the worldwide fallout (NCRP 

1990). Local fallout particles are significantly bigger (100 μm to several millimeters) (NCRP 1990). 

When deposited on skin or clothing, they can produce high levels of localized exposure, primarily from 

beta or alpha particles (NRDL 1968; NCRP 1990). 

Hot particle exposure is not easily identified because, in general, the dosimeter response cannot be 

used to distinguish between a hot particle exposure from a distributed radiation field exposure [11]. 

Because workers were normally monitored with hand-held survey instruments when leaving a 

radiation area, it was most likely that hot particles would be detected during this monitoring process. 

If such particles were detected, the worker would be asked to go through a decontamination 

procedure, including removal of outer clothing and showering as necessary (Allen and Schoengold 

1995). However, it is not likely that such incidents would be entered in the dosimetry record unless 

the conditions were unusual (e.g. very high count rates).

Unless hot particles are detected shortly after deposition using survey instruments, they will be 

removed by normal washing or change of clothes, and their exposure will not be recorded. Without a 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 25 of 121

specific entry in the dosimetry record, the only evidence, if the hot particle exposure is high enough, 

could be subsequent formation of a lesion at the deposition site [12]. 

The dose reconstructor should be aware of the possibility of external exposure from hot particle 

deposition and document positive indications in the claimant’s dose or medical record that can be 

reasonably associated with hot particle deposition. Hot particle deposition issues should be 

addressed consistent with project guidance on assigning shallow dose.

6.3.1.5.2 Intentional Nonuse of Dosimeters

There is anecdotal evidence suggesting that workers might have intentionally failed to wear their 

dosimeters during specific operations that could have caused them to exceed administrative limits 

[13]. This was said to have been done to avoid administrative actions that, inter alia, could have had 

adverse financial impact such as loss of overtime pay. Although there are no documented instances 

of this occurring, the practice could have taken place. In such situations, cohort dosimetry would be 

of questionable value because other members of the cohort are likely to have taken the same action. 

Dose reconstructors are encouraged to be aware that intentional nonuse of dosimeters could have 

taken place, and any indication of such occurrences should be documented in the claimant’s file. It 

should be noted that, because the personnel security credential holder (identification badge) was 

required for entry into controlled areas, workers needed to have their dosimeters with them, limiting 

their ability to avoid wearing the dosimeter.

6.3.2

The dosimetry methods employed initially at NTS were adopted from techniques implemented at 

LANL from the beginning of the atmospheric weapons testing program (NRC 1989; Boone, Bennett, 

and Adams 1970). As the various atmospheric test series progressed, dosimeter configurations 

improved in relation to the radiation fields encountered, and the responsibility for dosimetry programs 

was delegated among the participating agencies and military organizations [14]. In 1955, the site 

contractor, REECo, assumed responsibility for most onsite radiological safety functions; this included 

a site-wide service based on the use of film dosimetry for photons, betas, and neutrons. These 

methods evolved and eventually gave way to other methods including TLD and nuclear track 

detection or track etch detectors (TEDs) (Allen and Schoengold 1995; DeMarre 2002). From the 

beginning, PICs were used if necessary to augment the passive dosimeters issued to NTS workers 

and visitors. More detail on the Dosimetry Technology associated with the monitoring methods used 

at the NTS can be found in Attachment A.

Individual Monitoring Methods

6.3.2.1 Beta/Gamma Dosimeters

The film dosimeter issued for the first test series at the NTS in 1951 – Ranger – was a DuPont 552 

packet, which included two films: a Type 502 low-range component (0.05 to 10 R) and a Type 510 

high-range component (5 to 50 R). Brass/cadmium strips, 0.020-in. thick, provided symmetrical 

filtration on both sides of the film, with the remainder of the film open and unfiltered (Shipman et al. 

1951).

A modified dosimeter was used by both the Nevada Proving Ground (NPG) and Desert Rock RadSafe groups for the second NTS series in 1951 – Buster-Jangle (NRC 1989). The DuPont 553 

dosimeter included a third, higher range film component – Type 606 (10 to 300 R) (Kean 1951; 

Shipman et al. 1951; Storm 1951). NPG film packets were used in the Los Alamos brass-cadmium 

badge with 0.020-inch-thick brass and cadmium filters plus an open window. Desert-Rock film 

packets were in sealed, clear plastic envelopes, and probably had 0.020-inch-thick lead filters, as 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 26 of 121

were used a few months later at NPG during operation TUMBLER-SNAPPER (Kean 1951; Shipman 

et al. 1951; Storm 1951).

The film dosimeter used by the NPG personnel for Tumbler-Snapper was designated DuPont 558 

packet. With the exception of the first test, ABLE, it included lead filters, 0.028-in. thick, symmetrical 

on both sides; a Type 508 low-range component (0.01 to 6 R); and a Type 1290 high-range element 

(20 to 3000 R) (Brady and Nelson 1985).

The use of DuPont 559 dosimeters began in 1953 and continued to July 1960. It consisted of a Type 

502 low-range component (0.02 to 10 R), and a Type 606 high-range element (10 to 300 R). Lead 

filters, 0.028-in. thick, were used to provide symmetrical coverage on both sides, leaving an open area 

(Brady and Nelson 1985; Collison 1953). It was used essentially unchanged until HARDTACK II in 

1958, when Type 834 high range film (5–800 R) replaced the Type 606 film to provide and overlap 

with the Type 502 film.

The film badge dosimeter used from July 1960 through 1965 was a modification of the DuPont 559, 

designated DuPont type 301-4. It consisted of a type 508 low-range component (0.03 to 10 R) and a 

type 834 high-range component (10 to 1,000 R) wrapped with a 28-mil (0.028-in.)-thick lead strip 

covering an area 0.5 in. by 1 in. on each side. The packet was in a 4-mil (0.04-in.)-thick plastic bag 

sealed with colored tape to indicate the month of validity. The bag was clipped to the security badge, 

and all personnel entering NTS wore this dosimeter (DeMarre 2002). 

In 1966, NTS began using a combination personnel dosimeter and security credential holder (Brady 

and Iverson 1968) to provide the increased personnel dosimetry capability necessary to meet 

radiation exposure problems associated with nuclear rocket testing and underground nuclear 

detonations. The holder was designed to accommodate a DuPont type 556 film packet, a fast neutron 

packet [containing Kodak nuclear track emulsion, type A (NTA) film], an identification plate, criticality 

accident components, the security credential, and a snap-type clip. The complete package could 

measure beta, gamma, X-ray, thermal neutron, fast neutron, 

high-range gamma, and high-range neutron exposures.

In March 1971, when the use of DuPont film ended, NTS 

dosimetry operations converted to Kodak Type Ill film packets

(DeMarre 2002). This two-component packet contained lowrange (30 mR to 10 R) and high-range (10 to 800 R) films. The 

other components of the badge remained essentially the same.

The first routine use of TLDs at NTS began in 1970 (Boone, 

Bennett, and Adams 1970). Starting in February 1966, Pan Am 

used TLDs at the NRDS as part of the site effluent monitoring 

program (Figure 6-2) (PanAm 1967). These dosimeters 

contained a calcium-fluoride phosphor bound to a helically wound 

wire in an evacuated glass tube and were ideal for the intended 

purpose but unsuitable for personnel dosimetry (PanAm 1967). 

With the advent of DOE requirements to restrict personnel 

exposures to as low as reasonably achievable (ALARA) and with 

emphasis on accurate dosimetry at low doses, REECo 

Environmental Sciences Department personnel began evaluating 

TLD systems and neutron dosimeters in the early 1980s to 

replace the film badge and neutron TLD (DeMarre 2002). 

Figure 6-2. Security badge for 

NRDS TLD.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 27 of 121

After evaluating several dosimetry systems, the Environmental Sciences Department determined that 

the Panasonic 802 TLD and the neutron TED were the best combination for NTS exposure conditions. 

These were put into use January 1, 1987 (DeMarre 2002). The security credential holder was 

redesigned to accommodate both dosimeters (Figures 6-2, 6-3, and 6-4).

Figure 6-3. REECo badge holder (REECo 1990).

Figure 6-4. REECo security badge holder with Panasonic 

UD-802 dosimeter. 

The four-element Panasonic UD-802 TLD was the primary dosimeter for routine use issued to all 

monitored personnel until 2001, when it was replaced by the Panasonic 809 dosimeter, which also 

contained four elements (Table 6-1). The Panasonic UD-807 TLD is used to monitor personnel 

working in situations where the likelihood of exposure to an extremity is significantly greater than 

exposure to the whole body (Bechtel Nevada 2001). 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 28 of 121

In addition to film and TLD badges, self-reading pocket dosimeters were issued to persons entering 

controlled RADEX areas, which were controlled locations at which an exposure was usually expected 

(DeMarre 2002). The purpose of issuing PICs to persons entering a RADEX area was to provide an 

action alert. A high reading of a PIC triggered the action of collecting and processing the personnel 

dosimeter being used at the time. PIC results were not used unless the personnel dosimeter had 

somehow been compromised. The PIC result would have been included in a special investigation of 

the incident.

Although PIC results were only used as dose-of-record when film badge results were not available 

because of their typical over-response characteristics (Figure 6-1), PIC results should be evaluated 

carefully before including in the dose reconstruction.

6.3.2.1.1 Beta Dosimetry with Film Badges

Film response to beta radiation is accomplished with an unshielded portion of a film packet. 

Multielement film badges have an OW or unshielded portion that allows beta and photon radiation to 

reach the film. The beta dose is interpreted by removing the density attributable to a concomitant 

photon exposure as determined by the densities on those portions of the film under the various filters. 

The process is subject to large uncertainties, perhaps as much as a factor of 2 [15]. 

Prior to 1966, film dosimetry at NTS was performed with a bare film packet partially covered by a lead 

strip. The exposed film was interpreted by attributing the density under the lead filter to photons. The 

density in the unshielded portion of the film was assumed to be due to both photons and betas, and 

was typically not measured. In those instances in which the unshielded portion was read, the beta 

dose was determined by simply subtracting the density under the lead strip from that in the unshielded 

portion of the film to obtain a net density. This procedure did not accurately determine beta dose 

because it attributed the entire net density of the unshielded portion to beta dose, ignoring the lowenergy photon dose contribution (e.g., energies less than about 80 keV, of concern primarily to posttest drillback crews before 1965 and tunnel reentry crews in those situations when the tunnel was not 

vented before reentry), which did not penetrate the lead strip. It is estimated that as much as 25% of 

the total photon dose could have been missed as a result of this attenuation (Kathren 2004; Coryell 

and Sugarman 1951; Nelms and Cooper 1959). Therefore, the beta dose could have been 

overestimated. Given the spectra and mix of radionuclides encountered at NTS, dose estimates 

made in this fashion are likely to be within a factor of 2 of actual beta dose (Becker 1966, p. 102; 

Kathren and Larson 1969) and could slightly underestimate the overall photon dose. However, such 

dose estimates are likely to be unreliable and should be confirmed by other factors such as the 

beta:photon dose rate ratio obtained with monitoring instruments in the field, if such data are 

available. The value of a factor of 2 is an estimate of the range of uncertainty based on knowledge of 

the reported response characteristics of the dosimeters, and is presented for general information only. 

Multielement badges were introduced at NTS in 1966 (Brady and Iverson 1968; DeMarre 2002). A 

single high-Z metallic filter provided a more or less flat energy response for photons with energies 

above the uncertainty edge of the filter (about 50 keV). The density under the filter was used to 

assess the dose to photons above this cutoff energy and to evaluate the doses to photons with 

energies below the cutoff using densities under the other filters. Two filters – an OW and a low-Z

material (typically plastic such as Teflon) – were used to determine the beta dose. The response of 

the film under these filters was approximately the same for photons, but the low-Z plastic filter 

essentially removed all the betas. The reading from the low-Z filter, which was considered to be 

attributable to photons only, was subtracted from the density under the unshielded portion, which had 

a comparable response to photons and also was responsive to beta radiation. The analysis can be 

performed manually with a set of calibration curves made at different photon energies and a beta 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 29 of 121

calibration curve or with an algorithm developed from such a set of calibration curves. If it is 

necessary to make a beta dose estimate using the film badge results from shallow dose estimates 

from 1966 through 1986, dose reconstructors should double the reported value to ensure favorability 

to claimants and to account for uncertainties [15]. 

For external dose reconstruction, a positive indication of beta exposure in a dose record is considered 

to be due to betas with energies above 15 keV [16]. 

6.3.2.2 Neutron Dosimeters

A small fraction of the workers at NTS had potential for exposure to neutrons [17]. For workers with a 

possibility of neutron exposure, personnel neutron dosimeters were used to monitor exposure. 

Potential sources of neutron exposure at NTS were [17]: 

1. Direct production from a nuclear detonation

2. Spontaneous fission and subcritical multiplication in fissile materials (e.g., 235

U, 239Pu)

3. Isotopic sources such as initiators and calibration sources

4. Reactor testing

6.3.2.2.1 Film Dosimeters

From 1961 to 1971, nuclear track emulsions in the form of Kodak NTA film packets were used for 

personnel fast neutron dosimetry at NTS. The NTA film packet was incorporated in the newly 

designed combination personnel dosimeter and security credential along with a DuPont type 556 film 

packet for beta photon monitoring, an identification plate, criticality accident components, the security 

credential, and a snap-type clip (Brady and Iverson 1968). The NTA film packet consisted of a single 

thick-coated film of dental size wrapped in several layers of light-tight paper. More information on 

NTA film can be found in Attachment A, Section A.2.1.

From 1966 to 1986, every NTS badge had a cadmium strip that provided a thermal neutron-sensitive 

component that was only evaluated to provide verification that there had been no incidental neutron 

exposure (Brady and Iverson 1968). The density under a tantalum filter was compared to the density 

under the cadmium-tantalum filter to ensure that they were similar (equivalent densitometer reading of 

less than 80 mR). Cadmium readily absorbs thermal neutrons and subsequent to the absorption 

emits a photon of approximately 80 keV, which is of low enough energy to result in an over-response 

of the film badge compared to higher and lower photon energies. This energy is in the peak of the 

response curve, thus low level exposures to thermal neutrons were readily detected.

6.3.2.2.2 Thermoluminescent Albedo Dosimeters

The NTS albedo dosimeters adopted in 1979 were based on the design of Hankins (1977). They 

were used until 1987 and consisted of four TLD chips each of 6

LiF and 7LiF, placed in a cadmium box 

to suppress response to thermal neutrons. Albedo neutron dosimetry depends on the detection of 

low-energy neutrons reflected from the body (albedo neutrons) with a thermal neutron detector. 

Normally, a TLD with 6

LiF (TLD-600) is used to detect neutrons, while a companion 7LiF (TLD-700) 

detector that is insensitive to neutrons is used to subtract the photon/beta contribution.

Until January 1987, dosimeters were routinely exchanged each month for all individuals and on exit 

from a RADEX area if it was suspected that an individual received 100 mR or more of exposure. In 

addition, personnel entering RADEX areas were issued self-reading PICs. Measured exposures were 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 30 of 121

added to the yearly and quarterly accumulated exposures (REECo 1990; Allen and Schoengold 

1995). 

6.3.2.2.3 Track Etch Detectors

Gamma-insensitive Columbia Resin (CR)-39 TEDs were introduced for personnel neutron dosimetry 

at NTS in 1987. These were based on the dosimeter developed at LLNL (Hankins, Homann, and 

Westermark 1987; Hadlock et al. 1988). The CR-39 TED offered better fast neutron (>0.1 MeV) 

energy response characteristics for occupational monitoring than the TLD albedo dosimeter. The 

response curve is relatively flat, between 0.1 and 4 MeV (Attachment A, Figure A-5). Once a foil has 

been properly etched, acquiring information is a nondestructive process.

6.3.2.2.4 Panasonic 809

The Panasonic 809 replaced the Panasonic 802 dosimeter in 2001 (Bechtel Nevada 2001). The 809 

dosimeter contained three neutron-sensitive 6

Li2

10B4O7(Cu) TLD chips. This gives the dosimeter a 

neutron detection capability, so the use of the TED was discontinued. Photon compensation is 

provided by the neutron-insensitive 7

Li2

11B4O7(Cu) chip. The algorithm used to process the response 

of the four TLDs provides the neutron dose measurement. It is, in effect, an albedo detector.

6.3.2.3 Extremity Monitoring

Extremity dosimetry has been used at NTS to assess exposure to the finger, hand, forearms and even 

the head (on rare occasions) that might have occurred during operations in proximity to, or the manual 

manipulation of, radioactive material and radiation-emitting objects. Extremity monitoring might be 

required, for example, when radiation technicians were involved in handling post-test core samples.

The dosimeter (film or TLD) was worn in a position that was intended to represent the highest 

exposure to the extremity, usually on the inside of the wrist in the case of film, or on the finger near 

the finger tip. The extremity being monitored is normally identified in the dose record using the codes 

shown in ORAUT (2006b, Table A-2).

In 1957, extremity limits were set at 1,500 mrem/week (DeMarre 1993). The film pack used at that 

time was the Film badge-DuPont 559 film packet (Table 6-1). This continued until July 1960 when the 

use of DuPont film packet type 301-4 (Also called DuPont Type 556) was adopted (Table 6-1). In 

1964, the limit for extremities was set at 75 rem per quarter (DeMarre 1993). TLD finger rings for 

extremity monitoring were used beginning in July 1967 (DeMarre 1993). 

Although the regular use of TLD finger rings was documented in 1967 (DeMarre 1993), extremity 

monitoring with film and, later, TLD occurred on rare occasions prior to that time (DeMarre 2006c). 

ConRad 7

Li F and Teflon discs, 1.3 cm in diameter by 0.4 mm thick were used for measuring finger 

and hand exposures when personnel handled radioactive material in certain NTS operations (REECo 

1968). The 7

LiF and Teflon disks were protected from the light by inserting them in black plastic 

pouches. The pouches were affixed to the adhesive portion of “band-aids,”, which were attached to 

the fingers of the personnel being monitored. Extremity dose determined by the TLDs was not 

included in the routine dose reports. Rather, they were included with the bioassay data, and a card 

file of extremity doses was maintained.

The current extremity monitoring dosimeter is the single-element Panasonic UD-807 TLD shown in 

Figure 6-5. The Panasonic UD-807 TLD is similar to a single CaSO4:Tm element in the Panasonic 

UD-802 TLD. A single UD-807 TLD is sealed in a small, transparent, circular envelope embossed 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 31 of 121

with the TLD serial number. Except when removed for placement in an envelope-type holder, the 

TLD is kept in a processing holder. The holder has a serial number identical to the number embossed 

on the TLD element encapsulation. All personnel monitoring, exposure checks, and calibrations are 

performed with the TLD in the envelope-type holder, while all processing is performed with the TLD in 

the Panasonic holder.

Figure 6-5. Panasonic UD-807 extremity dosimeter (Allen and Schoengold 1995).

6.3.3

6.3.3.1 Photons

Calibration

In the early 1950s, film dosimeters were calibrated to free-air exposures using 226Ra and 60Co sources 

with source strengths traceable to the National Bureau of Standards (NBS), which is now the National 

Institute of Standards and Technology (NIST) (NRC 1989; Allen and Schoengold 1995). Exposures 

were quantified from inverse-square relationships over a fixed exposure interval at various distances 

and for exposures at fixed distances over various time intervals. Different approaches were used to 

compare dosimetry results, including processing badges exposed to calibration fields to badges 

exposed during test operations. In addition, collocated badges of different types and ionization 

chambers were exposed simultaneously during some test events to provide in-field calibration for 

realistic radiation spectra [18]. 

6.3.3.2 Neutrons

Neutron calibrations have been performed with isotopic sources, reactor beams, or accelerators. 

However, virtually all routine or practical radiation protection calibrations employ radioactive sources 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 32 of 121

[19]. The preferred neutron production mechanisms are either alpha-neutron reactions with beryllium 

or boron, or spontaneous fission (252Cf). On rare occasions, gamma-neutron reactions have been 

used, but the associated high-intensity gamma fields generally make them unacceptable for 

calibration of dosimeters and instruments that have any degree of gamma sensitivity.

Table 6-5 summarizes the properties of commonly used isotopic sources from the International 

Organization for Standardization (ISO 2001). Probably the first isotopic source to be used was 

226RaBe (Roberts 1937). However, its neutron production was accompanied by a high photon 

emission. As 239Pu, 238Pu, and 241Am became available, they became the alpha-emitting radionuclides 

of choice. Because these radionuclides have similar alpha energies, the resultant neutron energies 

are also similar and can be considered nearly identical for radiation protection purposes. In particular, 

210PoBe has been in use at NTS for several years (DeMarre 1993). Later 238Pu(n,α)Be sources were 

used (Allen and Schoengold 1995) 

Table 6-5. Characteristics of ISO reference radiation radionuclide

sources used for neutron dosimeter calibration. 

Source

Half-life

(yr)

Flux average energy 

(MeV)

AmBe-241 (α,n) 432.7 4.4

Cf-252 spontaneous fission 2.65 2.4

Cf-252 in 30-cm-diameter D2O sphere 2.65 0.54

Occupational neutron fields at the NTS are commonly due to fission-origin neutrons. Therefore, 252Cf 

is particularly attractive as a calibration source because it offers better inherent spectral simulation 

than the alpha-neutron sources. However, the unmoderated californium neutron spectrum is 

significantly harder than most reactor spectra. Beginning in the 1980s, spheres filled with deuterated 

water (D2O) came into common use as moderators to soften the spectrum of neutrons from the 252Cf 

source in the center (Schwartz and Eisenhauer 1980). 

For some neutron dosimeters such as TLD albedos, an alternative to source calibrations is the use of 

operational area calibrations. This involves exposure of dosimeters on appropriate phantoms in 

operational areas for extended periods (hours, days, or even weeks, depending on the ambient field 

intensities). The reading on the dosimeter when processed is then compared with a measurement 

made with a reference instrument. This relationship is used to establish a calibration for that 

particular area. However, the use of operational area calibrations implies work areas that have 

sufficient field intensities, so this method was not commonly used at NTS [20]. 

6.3.4

6.3.4.1 Gamma Dosimeters

Limit of Detection

6.3.4.1.1 Film

The National Research Council (NRC) evaluation of Film Badge Dosimetry in Atmospheric Tests

(NRC 1989) included evaluation of the minimum detection levels for film dosimetry used during the 

period from 1951 to 1962. Application of this concept to film dosimetry during atmospheric tests 

generally results in an MDL of approximately 40 mR, indicating that 95% of a series of exposures at 

40 mR would yield readings between 0 and 80 mR. Although lower values have been cited elsewhere 

(DeMarre 1993; Allen and Schoengold 1995), the NRC values are considered conservative, and they 

are presented in Table 6-1.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 33 of 121

6.3.4.1.2 TLD

Panasonic UD-802 TLDs have been tested to determine their lower limit of detection (LLD). The LLD 

is the minimum evaluated dose equivalent for which the readout value of a dosimeter is significantly 

different (at the 90% confidence level) from the mean readout of unirradiated dosimeters. LLD is a 

detection limit based on the standard deviation of background measurements and a 5% chance of 

reporting a false positive value.

DOE (1986a, Chapter 3) summarizes the procedure for determining the LLD for the DOELAP 

personnel dosimetry systems. It provides two alternative equations to calculate LLD. Although DOE 

(1986a) requires determination of the LLD for accredited dosimetry systems, no performance criteria 

are applied to these results.

The method used to determine the LLD for the NTS personnel TLD system is consistent with the 

method described in DOE (1986a). Use the following equations for determining the LLD:

[ ( ) ] [ ( ) ] ' 2 2

D1 p 0 += .12 75 o 1/1/1/ +−+ BSBHStL (6-2)

and

2 = pD StL 0 (6-3)

where:

LD1 = LD where the probability of reporting a false positive or false negative result is 5%.

LD2 = LD corresponding to a 5% chance of reporting a false positive.

Tp = distribution factor for n-1 degrees of freedom and a probability value of 0.95 = 1.68488.

So = [{∑(Xio – Ho)

2

}/(n-1)]1⁄2 = standard deviation of measurement

Ho = Ho’ = (Xio)/n = mean dose value

Xio = unirradiated dosimeter value background dose

N = number of dosimeters, number of measurements

B = bias for DOELAP categories

S = standard deviation for DOELAP categories

The LLD is primarily a function of the standard deviation in background dose rates. Table 6-6 lists 

selected NTS background dose equivalent rates. The values can vary over time. However, the 

temporal variation has been small since the end of atmospheric testing.

Table 6-6. Background dose equivalent rates (Allen and Schoengold 

1995). 

Location

Dose equivalent rate (mrem/d)

Shallow Eye Deep

NTS Area 12, Camp 0.375 0.311 0.283

NTS Area 6, Building 2 0.327 0.246 0.211

NTS Area 6, Building 50 0.255 0.182 0.150

NTS Area 23, Building 650 0.271 0.199 0.168

Test Range Complex, DOD Facility 0.249 0.193 0.169

Las Vegas, REECo Highland Building 0.181 0.139 0.121

Mean dose rate 0.276 0.212 0.184

Standard deviation 0.067 0.059 0.057
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 34 of 121

Table 6-7 lists detection limits for DOELAP categories determined during the 1990 accreditation 

process (Allen and Schoengold 1995). The LLD used at NTS as most appropriate for its operations is 

LD2, which is considered to be the most applicable value for determination of a dose likely to be higher 

than background levels and which increases the probability that a dose will be evaluated as positive. 

However, for dose reconstruction purposes, the major concern is reporting a false negative—reporting 

an exposure as zero when it really was greater than zero. Therefore, the LD1 values are 

Table 6-7. NTS Panasonic UD-802 detection limits for DOELAP categories (mrem/day). 

Category

Shallow dose Eye dose Deep dose

LD1 LD2 LD1 LD2 LD1 LD2

I. Low-energy photons–X-ray high dose N/Aa N/A N/A N/A 0.21 0.21

II. High-energy photons–high dose N/A N/A N/A N/A 0.21 0.21

IIIa. Low-energy photons–X-ray general 0.31 0.13 0.23 0.10 0.23 0.10

IIIb. Low-energy photons X-ray–plutonium environments 0.33 0.13 0.25 0.10 0.24 0.10

IV. High-energy photons 0.28 0.13 0.21 0.10 0.21 0.10

V. Beta particles–general 0.27 0.13 N/A N/A N/A N/A

a. N/A = not applicable.

considered to be appropriate for determination of MDL. Values of 0.3 mrem/d for deep dose and eye 

dose equivalents and 0.27 mrem/d for shallow dose should be used for NTS personnel TLD 

measurements. For a quarterly exchange period (90 d), the equivalent LLD values are 27 mrem for 

deep dose and eye dose equivalents, and 24.3 mrem for beta dose equivalent. Considering the 

conservatism of a 90-d exposure quarter, nominal values of 30 mrem for photon exposures and 25 

mrem for beta exposures are recommended for dose reconstruction (Table 6-1).

6.3.4.2 Neutron Dosimeters

Because of the energy dependence of the various personnel neutron dosimeters used at NTS 

(Attachment A, Section A.2), as well as other facilities throughout the DOE complex, the detection 

limits for neutron dosimeters are highly dependent on the operational neutron spectra encountered. 

The issue of operational neutron spectra is discussed in more detail in Section 6.3.5.3 and 

Attachment A.

6.3.4.2.1 NTA Film

NTA film has an effective energy threshold of about 0.5 MeV (Figure A-3). As a result, the limit of 

detection is directly proportional to the fraction of the neutron dose that is due to neutrons below 0.5

MeV (i.e., if 50% of the neutron dose is due to neutrons below 0.5, the limit of detection is twice as 

high as it would be for a PuBe calibration source).

NTA film dosimetry is relatively insensitive when compared with other methods. Under the best of 

conditions, with a fast neutron spectrum such as that from an n,α source or unmoderated 252Cf fission 

neutrons, a reasonable limit of detection is about 50 mrem (ICRU 2001). However, when the 

spectrum is degraded, the detection limit is increased significantly—by as much as a factor of 2 due to 

shielding and scatter from isotopic sources, and a factor of 5 for heavily shielded reactor operations 

(Section 6.3.5.3). As a result, the MDL recommended in Table 6-1 is 100 mrem.

6.3.4.2.2 TLD Albedo

TLD albedo dosimetry is inherently much more sensitive than NTA film dosimetry. However, it suffers 

from an adverse energy dependence, with the response decreasing seriously with increasing energy 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 35 of 121

(Figure A-4). Taking into account a bias of about a factor of 2 (Section 6.3.5.3), an MDL of 20 mrem 

is recommended in Table 6-1. 

6.3.4.2.3 Track Etch Detectors

Control foils processed with each TED processing are used as the basis for background subtraction 

for the TED system. The mean background dose equivalent to the control foils and the variance 

among the control foil dose equivalents are calculated for each processing. The background dose 

equivalent is determined for each processing because it can differ among foil sheets and within foils 

from the same sheet. Therefore, the TED background correction is not a single value, but a variable 

determined with each processing (Allen and Schoengold 1995; ICRU 2001). 

The method for determining the LLD for the TED system, with a variable background correction, was 

determined from Chapter 7 of NCRP Report 58, A Handbook of Radioactivity Measurements 

Procedures (NCRP 1978). The appropriate equations are:

LD1 = 2.71 + 4.65√VB (6-4)

and

LD2 = 2.32√VB (6-5)

where:

VB = background variance in units of dose equivalent squared (millirem squared).

The LD2 equation above was selected as the most appropriate LLD for the TED system. The selection 

reasons are the same as those discussed above for the TLD system.

The control foil variance (VB) increases with TED foil age and typically ranges between 5 and 25

mrem. Therefore, the TED system LLD is normally between 5 and 12 mrem. However, based on 

practical experience with these detectors, a value of 25 to 30 mrem is more realistic. That range of 

limits is reasonable for any TED issue period because it is primarily dependent on foil age and is only 

slightly dependent on the issue period. Considering the possibility of long-term variations and other 

contributing factors, potential energy-dependent under-response to moderated spectra, and the ratio 

of HP(10) to dose equivalent H, dose reconstructors should use a value of 50 mrem for the MDL 

(Table 6-1) [21]. 

6.3.5

The radiation production characteristics at NTS have been outlined in ORAUT (2004). The potential 

for external radiation exposures arises primarily from the fission and activation products handled at 

NTS (ORAUT 2004, Table 2-2). In addition, there is limited potential for neutron exposure from 

handling TRU radionuclides, isotopic sources, and reactor operations.

Workplace Radiation Fields

6.3.5.1 Photon Fields

The residual radiation field following detonation of a nuclear weapon consists of radiation from fission 

products, activation products, and unfissioned uranium or plutonium. During atmospheric testing of 

fission and fusion devices, differences in photon fields of residual radioactivity from detonations were 

observed. These differences are caused by the relative abundance of a few radionuclides that were 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 36 of 121

produced in each atmospheric test (Hicks 1981a to 1981i). For example, a low-altitude detonation of 

a fusion weapon induces large quantities of activation products emitting high-energy gamma rays that 

dominate the residual radiation spectrum for the first few days following the detonation. In contrast, a 

low-altitude detonation of a fission weapon produces large quantities of fission products that emit a 

wide range of photon energies. With either type of weapon, depending on the design, there can be a 

large amount of activity from the 239Np produced, which can dominate the spectrum for several days.

Although the residual radiation intensity depends on a number of factors that can vary from test to 

test, relatively few radionuclides, common to all tests, contribute to the major part of the photon 

spectrum. The relative abundance of each of these radionuclides determines the spectrum. In all 

cases, the photon field is from photons with energies between approximately 100 keV and 2 MeV. 

There is very little contribution from photons with energies less than 100 keV with the exception of 

scattering from large area sources. In those cases, the scattered radiation was determined to have 

an energy of approximately 75 keV and to have contributed as much as 10% of the overall photon 

spectrum (Kathren 2004). The special case of exposure to noble gases during post-test drilling can 

involve a significant contribution to photons with energies below 100 keV (Section 6.3.5.1.2).

For external dose reconstruction, if the conditions of exposure (work area, operation, etc.) are 

unknown, dose reconstructors should use the assumption favorable to claimants that photon energies 

are between 30 and 250 keV. If the exposure was due to fresh fallout (early reentry teams), it would 

be reasonable and still favorable to claimants to assume that 75% of the photon dose was from 

photons with energies above 250 keV (Kathren 2004; Coryell and Sugarman 1951; Nelms and 

Cooper 1959). If there is adequate documentation linking exposures to a particular NTS work area 

and/or operation, Table 6-9 presents guidance on a reasonable allocation of the recorded exposure or 

personal dose equivalent to the energy groups 30 to 250 keV and >250 keV. The information in this 

table is based on the radionuclide inventories presented in ORAUT (2004, Table 2-2). For more 

detail, see Attachment B. For the period from 1961 to 1966 when the multielement dosimeter was 

introduced, a contribution amounting to 25% of the total dose should be included in the range of 30 to 

250 keV to account for low-energy photons attenuated by the lead filter that covered a portion of the 

film (Kathren 2004). 

6.3.5.1.1 Nuclear and Ramjet Engine Tests

Nuclear rocket and ramjet engine tests were conducted on NTS in Area 25 and Area 26, about 80 mi 

northwest of Las Vegas, Nevada, from July 1959 through September 1969 (DOE 1995b). 

Development of the nuclear rocket engine began in 1955 under joint sponsorship of the Air Force and 

the AEC (DOE 1995b). The Los Alamos Scientific Laboratory (LASL) was responsible for developing 

reactor technology for the nuclear rocket engine (Project ROVER) while the Air Force was responsible 

for the non-nuclear portion of the project. In 1960, the AEC and the National Aeronautics and Space 

Administration formed the Space Nuclear Propulsion Office (changed later to the Space Nuclear 

Systems Office) to administer development of an operational nuclear rocket (NERVA, Nuclear Engine 

for Rocket Vehicle Application). Such a nuclear-powered rocket was visualized for use in space 

travel, given the crew could be adequately protected from radiation produced by the operating engine.

In a somewhat parallel program (Project PLUTO), LRL in 1957 began development of a nuclear 

ramjet engine. Because this was designed as an air-breathing engine, it was visualized as being 

restricted to relatively low altitudes. No engine tests were conducted during 1970 and 1971. The final 

related reactor test was of a "nuclear furnace" with a replaceable core in a reusable test bed designed 

to provide an inexpensive approach to testing advanced fuels in full-scale reactor environments. The 

nuclear furnace was successfully tested in June and July 1972. All nuclear rocket engines and 

systems tests were terminated in January 1973 (DOE 1995c). 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 37 of 121

During operation, the nuclear and ramjet assemblies were very well shielded with essentially no 

potential for external exposure of personnel. However, the maintenance that was required between 

runs often required hands-on manipulation that could result in significant exposure to the photon fields 

from fresh fission products (NRDL 1968).

6.3.5.1.2 Noble Gas Exposure

Exposure to significant concentrations of noble gases - isotopes of krypton and xenon - was possible 

following an underground test (LRL 1964; LANL 1994). While such exposure was not of concern from 

atmospheric tests because of rapid dispersion, certain post-test operations following an underground 

test carried the risk of relatively sudden exposure to large concentrations of such gases (LANL 1994). 

These exposures could generally occur in two ways.

Post-test drilling operations were often conducted as soon after detonation as possible. As soon as 

the cavity had collapsed and the crater was created, the drill rigs were brought in to begin the process 

of core sampling. The rig basically consisted of the drill platform, with a structure to shelter the drillers 

at some distance from the drill rod entry point, and a separate platform called the monkey board 

several feet above the platform. As the drilling progressed, new sections of drill rod were added with 

workers on the monkey board. Drill mud was circulated down the drill hole for lubrication. When the 

drill rod reached the cavity, circulation was lost as the mud entered the cavity rather than returning up 

the drill hole. This allowed the gases trapped in the cavity to escape up the drill hole, with the 

potential for significant exposure to workers on the platform and monkey board. The magnitude of the 

exposure would decrease with time following the detonation in proportion to the radioactive decay of 

the volatile radionuclides released (noble gases, halogens, etc.).

While the iodine that escaped in this way is primarily considered an internal exposure hazard, the 

noble gases are of concern from an external exposure standpoint, enveloping workers in a cloud, but

not having significant potential for intake (LANL 1994). The primary fission product noble gas 

radionuclides are summarized in Table 6-8. Of these, 133,133mXe are the nuclides of primary concern. 

Their fission yields were large enough that they were produced in significant quantities, while their 

half-lives were long enough that they would still be present in significant quantities several days after 

the test, but not so long as to significantly reduce their specific activity, compared with the much 

longer lived 85Kr.

Table 6-8. Primary iodines and fission product noble gases with half-lives greater than 1 hr (for yields 

≥ 0.01 photon or beta per disintegration). 

Nuclide Half-life

Photon fractions Average maximum 

beta energy (keV)

Average beta energy 

<30 keV 30–250 keV >250 keV (keV)

Kr-85 10.7 yr 0.0 0.0 1.000 685 251

Kr-85m 4.5 hr 0.0 0.84 0.16 839 290

Kr-87 76.3 min 0.0 0.0 1.00 3,119 1,333

Kr-88 2.84 hr 0.015 0.235 0.75 968 362

I-131 8.02 d 0.041 0.036 0.923 578 182

I-132 2.295 hr 0.002 0.186 0.812 1,320 485

I-133 20.8 hr 0.009 0.001 0.990 1,142 367

I-134 52.5 min 0.20 0.30 0.950 1,630 610

I-135 6.57 hr 0.003 0.017 0.980 944 363

Xe-133 5.24 d 0.0 1.00 0.0 346 100

Xe-133m 2.19 d 0.0 1.00 0.0 - -

Xe-135 9.14 hr 0.0 0.97 0.03 890 301
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 38 of 121

The potential for noble gas exposure during post-test drilling was primarily limited to radiation 

technicians, drillers, and roughnecks. Such exposures were drastically reduced with the introduction 

of blow-out preventers in 1964 (LRL 1964, 1966). 

Reentry after tunnel tests also had the potential for noble gas exposure to reentering miners and other 

crew members. Ventilation was usually introduced into tunnels following the tests to reduce the 

potential for such exposure prior to reentry. However, if the ventilation was not working, or had not 

been provided at all, the potential existed. The potential for noble gas exposure during tunnel reentries was primarily limited to radiation technicians and miners [22]. 

6.3.5.2 Beta Particle Fields

This section applies only to whole-body external doses from beta radiation as could be encountered 

from a fallout field or residual fission product activity at a work location. It does not apply to situations 

in which beta radiation is deposited directly on the skin or clothing, but rather concerns in-air doses to 

which a worker could have been exposed. Although there is a large and highly useful body of 

literature pertaining to skin doses from deposition of beta-emitting contamination on the skin, less 

information is available on external (i.e., in-air) beta doses from fallout fields or residual fission product 

activity.

There are about 250 individual fission-produced radionuclides, most of which are beta particle 

emitters.2

In contrast to gamma rays and neutrons, whose attenuation in matter is exponential, beta rays have a 

finite range in matter determined by the energy of the beta particle. Because beta particles emitted by 

radioactive species are not monoenergetic, the range is usually specified in terms of the maximum 

energy of the beta particle spectrum. For the beta produced by the decay of fission and activation 

products, the maximum energy typically does not exceed 3 MeV,

 Many of these beta-emitting fission product radionuclides have very short half-lives – on 

the order of a few seconds or less – and quickly decay into other nuclides. Others decay over longer 

periods. Therefore, the composition and beta energy spectrum of fallout changes over time. In 

addition to beta-emitting fission products, activation products are produced by a nuclear detonation, 

although their contribution to overall beta dose is relatively small [23]. Unlike photons, which are 

emitted with discrete quantum energies, beta emission is characterized by a distribution of energies 

ranging from zero to a maximum value that is commonly used to characterize the spectrum. The 

average energy of a beta spectrum is typically about one-third of the maximum energy, and the total 

energy produced by beta particles from fission products is essentially the same as that from fission 

product gamma rays, namely about 7 MeV per fission (Glasstone and Dolan 1977). 

3

For external dose reconstruction, a positive indication of beta exposure in a dose record is considered 

to be due to betas with energies greater than 15 keV.

and the range of a 3-MeV particle 

in air is approximately 36 ft. Therefore, an individual at a distance greater than 36 ft from a fallout 

field would not receive an external dose from beta radiation associated with the decay of 

radionuclides produced by fission and fission-produced activation products. Similarly, an individual 

exposed to beta particles with energies below 70 keV would receive no beta dose to the skin because 

beta particles with energies below 70 keV have insufficient energy to penetrate the cornified outer 

layer of the skin. More details related to beta particle ranges and beta-photon ratios can be found in 

Attachment A, Section A.4, and Attachment C. 

 

2 Although the term beta particle can refer to both positron and negatron emission from the nucleus of an excited atom, as 

applied to fission-produced radionuclides it refers only to negatron (i.e., electron) emission because no known fission 

products emit positrons.

3 The few exceptions are fission products with very short half-lives, which can be ignored for all practical purposes.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 39 of 121

6.3.5.3 Neutron Fields

Table 6-9 lists areas and operations where neutron exposure could have occurred at NTS. From 

1966 to 1986, fast neutron monitoring was conducted using NTA film. Personnel who did not work in 

the areas listed in Table 6-9, or were not directly involved in operations during the periods indicated 

Table 6-9. Areas and operations where neutron exposure was possible.

a (Allen and Schoengold 

1995) 

Area Operation Neutron sources Beginning Final

5 Low-level waste site TRU waste 1970 Present

6 Nuclear device assembly Fission neutrons 1951 1992

25 NRDS and BREN tower calibrations 

and operation

Fission neutrons & neutron 

sources Cf-252, PuBe, AmBe

1966 1973

26 PLUTO Reactor (nuclear-powered 

ramjet engine) 

Fission neutrons & neutron 

sources Cf-252, PuBe, AmBe 

Mid 1960s Late 1960s

27 Nuclear explosive assembly using 

special nuclear material

Fission neutrons & neutron 

sources Cf-252, PuBe, AmBe

1958 1975

Various Down-hole well logging PuBe-238 or Cf-252 isotopic 

sources

1951 Present

Various Neutron detection instrument 

calibration facilities

PuBe-238 or Cf-252 isotopic 

sources

1955 Present

a. Source: DeMarre (2006d). 

were not issued personal neutron dosimeters. However, there was a thermal-neutron-sensitive 

component in every NTS film badge to record any indication of neutron exposure.

6.3.5.3.1 Weapons-Related Neutron Fields

Exposure of NTS workers to neutrons from nuclear detonations, while a theoretical possibility, was for 

all practical purposes nonexistent [24]. As shown in Attachment D, if an individual was more than 6 

km from a detonation site, the neutron dose would have been less than 1 mrem. Strict measures 

were taken to ensure that personnel were not exposed to the prompt radiation from the detonation 

and, at the locations where personnel could have been exposed to prompt neutrons from the blast, air 

attenuation and similar attenuation mechanisms would have reduced the energy and the fluence of 

the neutrons to negligible levels. The closest workers (test personnel) were at CP-2, and later CP-1 in 

Area 6 (DeMarre 2006c). Therefore, dose reconstructors should ignore neutron exposure from this 

source unless there is evidence that the claimant was within 6 km of one or more of the atmospheric 

test detonation points at time of detonation. Note that CP-1 and CP-2 locations were more than 6 km 

from the atmospheric test locations.

Neutron exposures were possible in the vicinity of test shapes or other significant quantities of fissile

materials.. Plutonium pits that are not associated with high explosives are referred to as “bare pits,” 

although all pits are sealed or encapsulated (Shipler 2004). Assembly and disassembly operations 

comprise the only times workers have been exposed to neutrons emanating from bare pits. 

Maximum radiation dose rates occur when workers handled bare pits. Only a few individuals were 

associated with final assembly, arming, and firing of test weapons. Assembly operations were the 

only times that workers, in proximity with the weapons components, were exposed to neutrons 

emanating from bare pits. The operations often involve direct hands-on manipulation where the 

distance from the surface of the pit to the dosimeter is approximately 30 cm. Lead aprons or other 

shielding has been used to reduce photon dose rates. In assembly or disassembly operations, where 

high explosives (HE) or other materials surround the pit, photon and neutron dose rates decrease 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 40 of 121

significantly, although photon dose rates decrease more rapidly with increased shielding. Most 

workers involved in final assembly, arming, and firing of test weapons would have been national 

laboratory employees and traveled to NTS [25]. 

The workplace neutron fields for specific types of nuclear weapons components are classified. 

Unclassified information on neutron spectra from nuclear weapons components is not available, but 

there are two sources, both with significant components above 2 MeV. Before about 1960, nuclear 

weapons could have contained 210PoBe or 238PuBe initiators (DOE 1997) with a higher neutron dose 

component relative to the measured photon dose. However, these were not used after 1960.

For exposures after 1960, if neutron dose information is not specifically available for those involved 

with final assembly and arming operations, photon exposure records, together with neutron-to-photon 

dose ratios can be used. The neutron-to-photon ratios can be derived from the experience at the 

Pantex Plant, where weapons assembly operations were conducted. Analysis of dose records for 

each Pantex worker with a positive neutron dose greater than 50 mrem for the period from 1993 to 

2003 yields a geometric mean of 0.81 and GSD of 1.51. An upper 95th-percentile value of 1.6 should 

be used for the neutron-to-photon dose ratio (ORAUT 2006c).

Assuming that 100% of the neutron doses were delivered by neutrons in the 0.1–2-MeV range is 

favorable to claimants. Although there are more penetrating neutrons with higher energies at NTS, 

the POC for deeper organs, such as the liver, is much larger, in the 0.1- 2 MeV-energy range than any 

other energy group, thus offsetting the higher dose at depth for the more energetic energy ranges.

6.3.5.3.2 Isotopic Neutron Sources

A more significant potential source of neutron exposure was from isotopic neutron sources such as 

238PuBe or 252Cf. These sources were used in specific activities such as instrument calibration and 

well logging. Only a few highly trained and specialized individuals, however, had access to such 

sources [26]. 

6.3.5.3.3 Reactor Operations

The final source of potential neutron exposure was reactor test operations. These occurred in specific 

areas (Areas 25 and 26) designated for that purpose (Allen and Schoengold 1995). The number of 

individuals with the potential for neutron exposure was relatively small. The potential for significant 

neutron exposure was further mitigated by the fact that simultaneous gamma exposures were much 

greater and thus likely to be the governing factor for exposure control (Allen and Schoengold 1995). 

Neutrons were produced only during reactor operation, and personnel were always at locations 

remote from the reactors. Therefore, neutron exposures were low or negligible during reactor 

operations or test periods. Additional information on reactor operations is available in ORAUT (2004). 

6.3.5.3.4 Dosimetric Characteristics of Neutron Fields

The dose from neutrons is a function of neutron fluence and neutron energy. Fewer “fast neutrons” 

(i.e., neutrons with kinetic energies exceeding several hundred electron volts) are required to produce 

a given level of dose in comparison with slow or moderated neutrons. This is reflected in the use of 

quality factors or neutron weighting factors applied to the absorbed dose to arrive at dose equivalent. 

These energy-dependent factors have changed in the last 50 yr as a result of new information on the 

relative biological effectiveness (RBE) of neutrons. Table 6-10 summarizes historical changes in the 

factors used in the United States to adjust measured absorbed doses for the higher radiological 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 41 of 121

impact associated with neutron exposures. They are shown in the neutron energy groups used in 

dose reconstruction.

Table 6-10. RBE, quality factors, or weighting factors for neutrons. 

Neutron 

energy

(MeV)

Dosimetry 

guidelinea 

RBE

Quality 

factorsb

Average quality 

factors used at NTS

Neutron weighting 

factorc wr

Factor to be applied 

to NTS neutron dose

2.5E10-8 3 2

2.35 5 2 

1E10-7

10

2

1E10-6 2

1E10-5 2

1E10-4 2

1E10-3 2

1E10-2 2.5 5.38 10 2

1E10-1 7.5

5E10-1 11 10.49 20 2 

1 11

2 10

7.56 10 1.35

2.5 9

5 8

7 7

10 6.5

14 7.5

20 8

40 7 Not applicable 5 Not applicable

60 5.5

a. Trilateral meeting in 1949 radiation protection guidelines (Fix, Gilbert, and Baumgartner 1994).

b. Recommendations of NCRP Report 38 (NCRP 1971).

c. ICRP Publication 60 (ICRP 1991). 

The RBE was used until 1971, after which quality factors were used. The ICRU introduced the 

radiation weighting factor wr in 1990 as part of its definition of HP(10). Although wr has not been 

adopted in U.S. regulations, it is necessary to convert from neutron doses obtained using NCRP 

Report 38 quality factors to HP(10) (NCRP 1971). Project guidance provides conversions from 

radiation weighting factors for respective Interactive RadioEpidemiological Program (IREP) input 

neutron energy ranges.

6.4 DOSE RECONSTRUCTION RECOMMENDATIONS

6.4.1

Photon radiation dose reconstruction for NTS workers requires consideration of several parameters of 

dose reconstruction as follows:

Photon Radiation

• Determination of additional dose to unmonitored workers before the routine assignment of 

personnel dosimeters to all workers beginning in April 1957

• Adjustments to the reported DOE photon dose associated with the dosimeter response and 

radiation field characteristics

• Determination of the missed dose for monitored workers for low dose results [less than MDL of 

the personnel dosimeter]
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 42 of 121

• Determination of the IREP input photon radiation energy fraction

• Determination of the appropriate radiation quantities and exposure geometries to be used to 

determine the target organ dose

6.4.1.1 Unmonitored External Radiation Photon Dose

Assignment of dosimeters to all persons at NTS began in April 1957 and ended in December 1992

[27]. After 1992, only persons identified as having the potential for exposure were issued dosimeters; 

however, any person who requested a dosimeter would be assigned one. As noted above, all 

significantly exposed persons are expected to have been monitored (Allen and Schoengold 1995). 

For potentially exposed unmonitored workers prior to 1958, an estimated dose favorable to the 

claimant can be assigned based on the 50% dose of the monitored workers noted in the last column 

of Table 6-11. This provides an estimate favorable to claimants of the potential unmonitored dose, 

particularly if assigned for several years of employment. This dose can be adjusted during dose 

reconstruction to reflect the actual annual employment of the exposed person as verified by DOL. 

Most workers were not continuously employed at this site.

A 50th and 95th percentile dose has been calculated for the badge exchanges of persons at the NTS 

and is presented in Table 6-11. For an unmonitored worker with verified employment prior to 1957 

the 50th percentile dose and missed dose for each month employed during the years evaluated 

should be assigned. Except for the 4th plutonium dispersal event (1/18/1956), there was no device 

testing during 1954 and 1956. Table 6-12 used in developing Table 6-11 is based on a query of the 

NTS historical dosimetry (1945-1983) database (DeMarre 2006e). The values include civilian workers 

and visitors and the military. An extensive review of test series measured doses may be viewed at 

http://www.dtra.mil/rd/programs/nuclear_personnel/atr.cfm.

Table 6-11. NTS calculated dose to apply as co-worker dose.

a

Year

Coworker dose (rem)

50% dose 95% dose

1945 0.001 4.251

1946 0.147 0.856

1947 0.036 0.499

1951 0.058 0.941

1952 0.106 0.665

1953 0.048 3.186

1954 0.001 0.172

1955 0.014 1.173

1956 0.026 0.323

1957 0.001 0.481

a. Summaries of external dose measurements were provided as a dose histogram. Dose values 

for the 50th and 95th percentile were calculated by finding the numerical position of each 

percentile value within the histogram to determine the correct bin. Then extrapolate to the point 

within the bin to calculate the dose. The method used to determine the 50th percentile and 95th 

percentile are inherently favorable to the claimant. It is likely the doses represented in each bin 

of the histogram would have been skewed towards the lower end of the bin (i.e. lognormally 

distributed). The analysis assumed that the dose points were equally spaced between the lower 

and upper limits of the bin which should have resulted in a higher calculated dose. 

Example: In 1951 there were 21,086 recorded dose measurements. If the measurements were 

sorted from smallest to largest then the position that would represent the 95th percentile would 

be 20,032 (21,086 * 0.95). Knowing the position, the histogram bin is also determined, which in 

this case is 0.500 – 0.999 rem. Now extrapolate to position 20,032 within that bin. The position 

number representing the bottom of the bin is 19,987 which means the 95th percentile value is in 

position 844.7 (20,032 – 19,987) for a bin that contains a total of 965 measurements. 

Extrapolation results in a dose of 0.941 rem (844.7/965) * (0.999 – 0.5) + 0.5.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 43 of 121

6.4.1.2 Adjustments to Recorded External Photon Radiation Dose

Table 6-1 provides bias and uncertainty factors determined for the respective test series. The factors 

are dependent upon the radiation field specifics of the test series and the dosimeters in use. These 

values likely provide the most reasonable option to adjust the NTS claimant reported dose. The

annual reported penetrating dose for each of the years of employment is divided by the bias (B) 

factors noted in Table 6-1 to arrive at the annual estimated exposure. The bias should be applied to 

both the reported and the missed dose, when applying the bias increases the dose being assigned.

6.4.1.3 Missed Recorded External Radiation Dose

Missed dose occurs when the dose of record is less than the MDL (i.e., considered to be zero) 

because the dosimeter response was less than the MDL or there is no dose of record for an assigned 

badge for a monitoring period. This kind of missed dose is most important for earlier years when 

MDLs were higher and dosimeter exchange was more frequent. Dose reconstructors should follow

NIOSH (2006) guidance and use the MDL data in Table 6-1 to calculate the missed photon dose as 

follows: 

Missed photon dose can be assigned based on the MDL/2 method and project guidance 

regarding reported dose under the MDL/2, and the number of exchange periods (NIOSH 2006) 

given in Table 6-1 for the dosimetry systems or the dose of record provided by DOE. For 

many workers, the number of dosimeter exchanges will be greater than the routine monthly 

exchange. The dose reconstructor should evaluate the exchange cycles and assign the 

appropriate number of zero cycles or default values in Table 6-1 as applicable.

6.4.1.4 IREP Input Photon Radiation Energy Fractions

Attachment B presents an inventory of the radionuclides encountered at NTS by area and operation. 

This information is used as a basis for estimating the IREP input fraction of the dose in each of the 

three energy groups – less than 30 keV, 30 to 250 keV, and greater than 250 keV. Default values are 

100% 30 to 250 keV for efficiently processing a claim or 25% 30 to 250 keV and 75% greater than 

250 keV. The results are presented in Table 6-13.

6.4.1.5 Determination of Radiation Quantities and Exposure Geometry

As noted in Attachments A, B, and C, the conversion coefficients for exposure and personal dose 

equivalent vary widely within each of the three energy groups – a factor of more than 10 for <30 eV 

(with a 10-keV cutoff), as much as a factor of 4 for the 30-to-250-keV group, and up to a factor of 10 

above 250 keV (4,000-keV upper cutoff). The recommended dose quantities to be used in selecting 

dose conversion factors from the External Dosimetry Implementation Guide (NIOSH 2006) are:

• 1951–1986: Exposure (R) to Organ Dose(HT) 

• 1987–Present: Deep Dose Equivalent [HP(10)] to Organ Dose (HT)

The anterior-posterior geometry should generally be selected because higher doses are typically 

recorded while workers are directly working with and facing sources of radiation.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 44 of 120

Table 6-12. NTS associated measured doses prior to universal badginga

Dose in rem

Lower 0 0.001 0.050 0.1 0.15 0.2 0.25 0.3 0.4 0.5 1.0 1.5 2.0 2.5 5.0 7.5

Higher 0 <0.05 <0.1 <0.15 <0.2 <0.25 <0.3 <0.4 <0.5 <1.0 <1.5 <2.0 <2.5 <5.0 <7.5 <10 >10

Year Number of personsb

1945c 776 8 23 31 14 24 14 23 17 94 87 22 25 157 2 10 7

1946 16 12 23 24 21 5 7 9 15 14 2 0 0 0 0 0 0

1951 6,849 2,730 5,851 1,735 537 421 382 457 225 956 311 137 49 444 2 0 0

1952 978 4,972 1,569 3,970 319 225 134 2,781 109 348 228 125 98 104 7 2 2

1953b 18,746 6,351 3,448 2,134 951 1,219 592 1,295 708 3,723 2,856 202 4,305 3,266 100 9 18

1954b 622 205 102 32 31 13 5 5 1 5 4 0 0 1 0 0 0

1955b 16,122 4,652 1,922 3,094 595 352 325 906 1,326 3,187 1,490 407 102 292 60 2 10

1956 254 471 114 56 25 16 8 13 7 18 7 3 0 1 0 0 4

1957 42,999 9,617 9,046 5,559 2,615 2,094 1,249 1,868 1,205 2,727 676 183 98 77 9 3 4

a. Results of query for NTS persons from NTS historical dosimetry (1945–1983) database (DeMarre 2006e).

b. The number of individuals is inflated for 1953-1955 because persons were counted by account number and could have more than one account number due to multiple IDs (NTS 

number, employee number, social security number). Dose reconstruction records are also counted (provided by DTRA for military participants) and there are visitor/vendors who 

were included as well. This would make the totals higher than just film badges alone.

c. No testing occurred at NTS in 1945 and 1946, however a continental atmospheric test occurred in New Mexico as Operation Trinity. Since some of the same individuals were 

potentially present during later tests, the coworker data for them has been included for completeness.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 45 of 121

Table 6-13. NTS work area and operation-dependent photon fractions.a

Operation Area

Assigned fraction

<30 keV 30-250 keV >250 keV

Drillback operations 1–10 and 18-20 0.03 0.50 0.47

Reentry and mineback operations 1, 12, 15, and 16 0.04 0.38 0.58

Routine tunnel operations 1, 12, 15, and 0.00 0.22 0.78

Decontamination facility 6 0.14 0.43 0.44

Treatability test facility 25 0.12 0.46 0.42

Atmospheric safety test areas 5 and 11 None indicated

Atmospheric weapons test areas 1–5, 7–11, and 18 0.10 0.45 0.45

Low-level waste site 3 0.31 0.40 0.29

Low-level waste site 5 0.20 0.45 0.34

Radiation instrument calibration 6 and 23 0.36 0.40 0.24

Radiograph operations 23 0.00 0.23 0.76

Well logging operations 1–10 and 18–20 0.32 0.38 0.30

Nuclear explosive/device assembly 6 and 27 0.57 0.43 0.00

Nuclear rocket development 25 and 26 0.02 0.43 0.55

Radioactive source storage 6 and 23 0.26 0.41 0.33

Radiochemistry and counting laboratories 6 and 23 0.09 0.40 0.51

a. See Attachment B for derivation of partition fraction. 

6.4.1.6 Correction Factors for External Environmental Dose

External environmental exposures, as recorded by the dosimeter, have geometries that are different 

from those used for dosimeter calibrations. The sources are highly extended, and isotropic in contrast 

with point sources placed relatively near the dosimeter for calibration. The dose assessment for 

external environmental exposures requires that attention be given to 1) angular or directional 

dependence of the dosimeter, 2) angular or directional dependence of the dosimetric quantity used for 

monitoring purposes, and 3) the dose of specific organs relative to the dosimeter. Environmental 

exposures, particularly at the NTS, are characterized by complex photon spectra due to the presence 

of a large number of fission and activation product radionuclides. In addition, these spectra are 

significantly modified by the attenuation of lower energy photons by several centimeters of intervening 

soil or several meters of intervening air.

The evaluation of geometric considerations and correction factors for assessment of external 

environmental exposures at the NTS has been addressed by Griffith (2006). The assessment 

included three distinct exposure geometries:

1. Exposure to ground surface contamination (infinite plane surface), characteristic of fresh 

fallout.

2. Exposure to soil contaminated to an infinite depth, characteristic of fallout that had been in 

place for several days to weeks, and had been “weathered in”.

3. Submersion in a semi-infinite cloud, characteristic of airborne radioactivity as might be 

encountered in a release following an underground test.

Data on the angular dependence of the dose conversion factors is taken from ICRU (1988, 1992, 

1998) and ICRP (1987). Angular dependence of the dosimeters is based on dosimeter design. 

Although specific information for NTS dosimeters was not available, data for dosimeter of similar 

design was taken from results of the International Collaborative Study of Cancer Risk among Workers 

in the Nuclear Industry (Cardis et al. 2005; Thierry-Chef et al. 2002).
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 46 of 121

Calculations of the correction factors for organ geometry relative to the dosimeter have been 

performed for exposure to 1) fallout from atmospheric tests (Hicks 1981c, 1981i), 2) radionuclides 

released by underground test leaks (Hicks 1981a), and 3) nuclear rocket and ramjet tests (Hicks 

1981a), for exposures from time of test to 50 years following the test. The radionuclide and geometry 

dependent dose conversion factors were taken from Eckerman et al. (1999). 

Results of these calculations (Griffith 2006) show that the correction factors for external exposure 

environmental radiation fields found at the NTS are not significantly different from unity for most 

organs, these values are less than 1. Given the low environmental external exposure rates at NTS, it 

appears that the new DCFs would not have a significant impact on the assigned environmental doses 

in comparison to the NIOSH dose estimates that are favorable to claimants.

6.4.2

6.4.2.1 Evaluation of Beta Exposure Geometries

Beta Radiation

In the event that personal dose records do not include estimates of beta dose, an estimate can be 

made from the associated photon (gamma + X-ray) exposure or dose equivalent using beta-photon 

ratios. Because the beta-photon ratios are highly variable and generally decrease with distance from 

the source, it is necessary to select an arbitrary but realistic geometry as a basis for the ratio 

estimates. Three common geometries associated with beta exposures are (1) standing on a 

contaminated surface, (2) immersion in contaminated clouds, or (3) exposure to discrete sources

(AECL 1982). 

To determine the best value of beta-photon to be applied to the relevant photon dose, it is necessary 

to clearly identify the claimant’s work requirements as they relate to the most likely exposure 

conditions. The beta-photon values that appear in this section are given in terms of sievert beta per 

sievert photons. However, with a quality factor equal to 1 for beta particles, 1 Sv = 1 Gy = 100 rad. 

For photons, 1 Sv = 1 Gy. Because 1 Gy = 100 rad = 114 R, within the uncertainties associated with 

establishing beta-photon ratios, the numerical value of the beta-photon ratio can be considered valid 

for the period covering the NTS operations [28]. 

If the claimant was working in connection with the nuclear testing or the nuclear rocket and ramjet 

development program or if they were primarily working in areas where fission and activation products 

from previous operations were a significant, if not the most significant, source of exposure, it is 

appropriate to use data for one of the geometries primarily associated with such exposure – surface 

contamination or plume/cloud immersion. If the claimant had work assignments at other areas of 

NTS, factors related to source inventories in those areas would be appropriate to use.

Detailed inventories of fission and activation products following an atmospheric nuclear detonation, 

release of contamination following detonation, or nuclear rocket and ramjet operation are also 

available (Hicks 1981a to 1981i, 1982, 1984). The development of beta-photon ratios for estimating 

external dose to the skin and eye from standing on contaminated surfaces or immersion in 

contaminated clouds is summarized in Attachment C. These values apply only to skin and eye. 

Separate beta/photon values for the gonads were not calculated (Kocher and Eckerman 1981; 

Eckerman et al. 1999) because the beta particles were not considered to contribute to the gonad dose 

beyond bremsstrahlung production in the soil and air (Eckerman 2006a).
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 47 of 121

6.4.2.2 Beta-Photon Ratios for Surface Contamination

External beta doses from standing on contaminated ground or other surfaces can be calculated by 

applying a beta-to-gamma dose ratio to an estimated upper bound gamma dose, which is determined 

from film badge data or dose reconstruction. The exposure from extended contaminated surfaces is 

discussed in more detail in Attachment C. Table C-1 presents time-dependent beta-photon ratios for 

fission and activation products from detonation or time of operation to 50 yr. Based on these data, the 

following beta-photon ratios are recommended if the exposure is presumed to have occurred for an 

extended contaminated surface (e.g., ground contamination from dispersal of fallout). The 

beta/photon ratios would be reduced by attenuation of clothing (anti-contamination clothing (e.g., 

coveralls), shoes, gloves, etc.), increased separation between the contaminated surface and the 

individual (e.g. working on an elevated structure), or any intervening material (surface coatings, 

floorings, etc.). The dose reconstructor should determine if the exposure resulted from the Energy 

Employee walking directly on the contaminated surface, or if the job site characteristics involved nonradioactive, intervening surfaces of materials.

Information in the claim files may support the appropriateness of using the calculated values listed 

below. However, to efficiently process a claim a value of 3 to 5 can be justified as reasonable, since 

many of the parameters listed above, cannot be quantified for the claim (i.e. distance of the worker 

from the source).

Elapsed time following the 

production event

Beta/photon

Sv/Sv

0 to 50 d 10

50 to 365 d 25

1 to 5 yr 60

>5 yr 25

6.4.2.3 Beta-Photon Ratios for Immersion in a Contaminated Cloud

Immersion in a contaminated fallout debris cloud was a less frequent circumstance than exposure to 

fallout after deposition on the ground or other surface [29]. However, exposure of workers to plumes 

of radioactive gas frequently occurred following releases from underground tests such as occurred 

during post-test drilling before gas blocking or during accidental venting (LRL 1966, LANL 1994). 

External beta doses from immersion in a contaminated cloud or plume can be calculated by applying 

a beta-to-gamma dose ratio to an estimated upper bound gamma dose, which is determined from film 

badge data or dose reconstruction [30]. 

The exposure from immersion is discussed in more detail in Attachment C. Table C-2 presents timedependent beta-photon ratios for fission and activation products from detonation or time of operation 

to 50 yr. Based on these data, the following beta-photon ratios are recommended if the exposure is 

presumed to have occurred for an extended contaminated surface (e.g., ground contamination from 

dispersal of fallout). Because the release of gases is primarily associated with weapons detonations 

at NTS, this geometry is recommended for drillback, reentry, and mineback operations.

Elapsed time following the 

production event

Beta/photon

Sv/Sv

0 to 50 d 1.5

50 to 365 d 5

1 to 5 yr 10

>5 yr 5
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 48 of 121

If the exposure is known to have resulted from immersion in a plume consisting only of noble gases 

(krypton and xenon) and halides (bromine and iodine), the following beta-photon ratios should be 

used (Attachment C, Figure C-4).

Elapsed time following the 

production event

Beta/photon

Sv/Sv

0 d 1

>100 d 100

6.4.2.4 Beta-Photon Ratios for Other NTS Operational Areas

If the presumed exposure was likely due to discrete sources or radioactive material with limited spatial 

distribution, application of beta-photon ratios associated with such exposure conditions can be used. 

The beta-photon ratio can be estimated from the radionuclide inventories known to be associated with 

a particular operation or work area. In some work situations, neither a contaminated ground surface 

nor immersion in a contaminated plume might adequately represent an exposure geometry. 

Attachment C contains detailed development of beta-photon ratios for a point-source geometry where 

the source is 1 m from the body surface. If sufficient information about the nature of a presumed 

exposure is available, more detailed calculations can be made on a case-by-case basis. Beta-photon 

ratios for the radionuclides in NTS operational areas are listed in Table 6-14. 

Table 6-14. Nominal beta-photon ratios for operational areas at NTS. 

Areas and operation Area

Period of 

operation

Beta HP(0.07)/photon HP(d)–Sv/Sv

Pure beta 

emitters

Surface 

contamination Point source

Decontamination facility 6 1971–2001 20. 90. Sr-89, Sr/Y-90

Treatability test facility 25 1990s 20. 100. Sr/Y-90

Atmospheric safety test areas 5 and 11 1956 –1957 0. 0. -

Low-level waste site 3 Mid 1960s–

present

25. 100. Sr/Y-90

Low-level waste site 5 1961–present 25. 65. Sr/Y-90

Radiation instrument 

calibration

6 and 23 1951–present 30. 3. Sr/Y-90

Radiograph operations 23 Unknown 15. 4. -

Well logging operations 1–10 and 

18–20

1963–1992 30. 5. - 

Nuclear explosive/device 

assembly

6 and 27 1951–present 0.1 0.1 - 

Radioactive source storage 6 and 23 1951–present 15. 25. Sr/Y-90

Radiochemistry and Counting 

Laboratories

6 and 23 1958–present 15. 80. Sr/Y-90

Atmospheric weapons test 

areas

1-5, 7-11, 

and 18

1951–1963 See Sections 6.4.2.2 and 6.4.2.3 Sr-89, Sr/Y-90

Drillback operations 1–10 and–20 1963–1992 See Sections 6.4.2.2 and 6.4.2.3 Sr-89, Sr/Y-90, Y-91

Reentry and Mineback 

Operations

1, 12, 15 and 

16

1957–1992 See Sections 6.4.2.2 and 6.4.2.3 Sr-89, Sr/Y-90, Y-91

Routine tunnel operations 1, 12, 15, 

and 16

1957–1992 See Sections 6.4.2.2 and 6.4.2.3 Sr-89, Sr/Y-90

Nuclear rocket development 25 and 26 1959–1973 See Sections 6.4.2.2 and 6.4.2.3 Sr-89, Sr/Y-90

6.4.2.5 Estimation of Beta-Photon Exposure

The simplest calculation is for a short-term exposure to beta and gamma radiation. The total beta 

plus gamma dose to the skin or lens of the eye is estimated as (Barss 2000): 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 49 of 121

Dose skin/lens = D(t)γ/ub/fall [Rβ/γ(x,t) × M(x,t) + 1] (6-6)

where:

D(t)γ/ub/fall = upper-bound gamma dose due to external exposure to fallout or other beta radiation 

field,

Rβ/γ(x,t) = β/γ ratio = bare skin or lens beta-to-gamma dose ratio at distance x and time t,

M(x,t) = combined modifying factor that accounts for differences from the simple case of 

standing on contaminated ground with bare skin (e.g., attenuation by clothing, 

position of the body, and location above or below deck on a ship or in an aircraft) 

[see Barss (2000), Attachment C, Table 12], and

D(t)γ/ub/total = upper-bound gamma dose from all sources.

Dose to skin is reduced by clothing, and Barss (2000) provides a method for determining the 

reduction as a function of clothing thickness. However, the corrections at a point 1 m above the 

contaminated surface reduce the beta exposure on the order of 20 to 30%. In view of the associated 

uncertainties, attenuation is also recommended consistent with project guidance on shallow dose 

(ORAUT 2005).

Because of the uncertainties associated with the application of beta-photon ratios to estimate potential 

beta exposures to NTS workers, dose reconstructors should only use these ratios when other relevant 

information, such as the optical density under the open area of a film badge, is not available. 

If the worker dosimetry record contains no other information, including portable survey instrument 

results, beta-photon ratio estimates may be used with caution, recognizing that the results can be 

considered semi-quantitative at best, with likely errors and uncertainties in the beta dose estimate 

easily exceeding a factor of two. Uncertainty factors of ten, considering the lack of information are not 

unreasonable [31]. 

6.4.2.6 Skin Contamination

For skin contamination (for example, from fallout or resuspended radioactive soil), the film badge 

gamma dose is a highly inaccurate indicator of skin dose, so beta-to-gamma ratios are not 

appropriate for such applications. However, the beta energy spectrum due to radioactive material on 

the surface of the skin can be determined as a function of time after detonation for each radionuclide. 

This allows beta doses to be directly calculated by using dose coefficients from Kocher and Eckerman 

(1987).

These dose coefficients are based on radionuclides deposited on or near the skin surface, and are 

nearly constant for average beta energies greater than 0.1 MeV. Therefore, if a measurement of the 

skin contamination levels was recorded, it is possible to estimate an average dose coefficient for skin. 

That dose coefficient is about 9 rem h−1 (beta plus gamma) per μCi/cm2 of skin (Barss 2000). This 

includes a 30 to 35% overestimate due to the potential presence of an external backscatter surface 

(e.g., a contaminated surface or tool) and a gamma contribution of about 5%. For contaminated 

gloves, a dose reduction factor of 0.5 is assumed. However, without recorded skin contamination 

levels, the skin dose is virtually impossible to determine.

Skin contamination should be based on measurements when they are available that provide 

information to guide estimates of skin contamination based on measurements expressed in terms of 

dose or exposure rate. The VARSKIN code (Durham 1992) can be used for additional calculations of 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 50 of 121

skin dose, presumably for source geometries where an assumption of uniform large-area 

contamination is inappropriate.

6.4.3

6.4.3.1 Introduction

Neutron Radiation

Project guidance has been developed for factors to be applied to neutron dose as follows. Figure 6-6 

shows the ratio of conversion coefficients for personal dose equivalent, HP(10, 0°) to conversion 

coefficients for dose equivalent, H (NCRP 1971). Using this figure, the last column of Table 6-10 lists 

factors to make this conversion for each of the four energy groups for dose reconstruction at NTS 

(none of the neutron-producing facilities generated a significant number of neutrons above 20 MeV). 

If the neutron spectra are not known, dose reconstructors should treat the exposure as having come 

from neutrons in the 0.1-to-2-MeV range, and should use a factor of 2 to convert from NCRP Report 

38-based values to HP(10). 

0

0.2

0.4

0.6

0.8

1

1.2

1.4

1.6

1.8

2

1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1.E+01 1.E+02

Neutron energy - MeV

Hp(10)/H

Figure 6-6. Ratio of fluence to dose conversion coefficients for personal dose 

equivalent HP(10, 0°) to conversion coefficients for dose equivalent H. 

The neutron spectra at NTS have been generally unmoderated and rich in fast neutrons, so much so 

that doses from thermal neutrons were most certainly trivial. However, in some operational situations, 

considerable scattering could have occurred, resulting in some softening of the spectra [32]. 

Attachment A, Section A.5 presents more detail on neutron spectral characteristics. Figure A-10 and 

Table A-4 indicate that the HP(10, 0) contribution from neutrons with energies less than 10 keV can be 

ignored for the purposes of NTS dose reconstruction. The primary HP(10) contributions fall in the 

energy ranges 100 keV to 2 MeV and 2 to 20 MeV. This information is summarized by energy band 

for a variety of spectra in Table 6-15 and NTS operational areas in Table 6-16. If the spectral 

characteristics of the operational neutron field are known, dose reconstructors can use this table to 

assign neutron dose fractions to the primary energy bands – 10 to 100 keV, 100 keV to 2 MeV and 2 

to 20 MeV. In the absence of such information, however, it is recommended that dose reconstructors 

make the assumption favorable to claimants that neutron energies are between 100 keV and 2 MeV.

Table A-4 also indicates that NTA film could underestimate doses by 55% or more because the 

detection threshold for the film is ≥ 0.5 MeV. Therefore, it is recommended that a bias for NTA 

measurements of 0.5 (Table 6-1) to account for potential dosimeter under-response in softer neutron 

fields. In addition, a correction for the ratio of dose equivalent H to personal dose equivalent HP(10)

conversion coefficients of 2 in the 0.1 to 2 MeV energy range (Table 6-10) should be applied. If the 

fractional neutron contribution in the 2 to 20 MeV range is well known (e.g. unmoderated isotopic 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 51 of 121

Table 6-15. Fraction of H=(10,0) by energy band for selected spectra. 

Spectral description

Neutron energy bands

<100 keV 100 keV–2 MeV 2 MeV–20 MeV

Cf-252 0.003 0.659 0.338

Cf-252 w/room scatter 0.017 0.691 0.292

Cf-252, 15 cm D2O 0.089 0.579 0.332

Cf-252, D2O w/room scatter 0.106 0.669 0.225

AmBe-241 0.001 0.261 0.738

AmBe w/ room scatter 0.010 0.366 0.624

PuBe at 1m 0.005 0.368 0.627

TRU Plant, Pu Repro. Plant, heavily shielded 0.080 0.734 0.186

Pu-238 O2 at 100 cm 0.021 0.821 0.158

Godiva spectrum 0.008 0.773 0.219

Table 6-16. Neutron field characteristics associated with NTS operational areas.

a

Area Operation Neutron sources

Neutron energy fractional distribution

Neutron energy–MeV Fraction

5 Low-level waste site (TRU waste) TRU waste 0.1–2.0 1.00

6 Nuclear device assembly Fission neutrons 0.1–2.0 1.00

25 NRDS and BREN tower calibrations 

and operation

Fission neutrons & neutron 

sources Cf-252, PuBe, 

AmBe

0.1–2.0 1.00

26 PLUTO Reactor (nuclear-powered 

ramjet engine) 

Fission Neutrons & neutron 

sources Cf-252, PuBe, 

AmBe 

0.1–2.0 1.00

27 Nuclear explosive assembly using 

special nuclear material

Fission neutrons & neutron 

sources Cf-252, PuBe, 

AmBe

0.1–2.0 1.00

Various Down-hole well logging PuBe-238 isotopic sources 0.1–2.0

2.0–20.

0.5

0.5

Various Down-hole well logging Cf-252 isotopic sources 0.1–2.0

2.0–20.

0.75

0.25

Various Neutron detection instrument 

calibration facilities

PuBe-238 isotopic sources 0.1–2.0

2.0–20.

0.5

0.5

Various Neutron detection instrument 

calibration facilities

Cf-252 isotopic sources 0.1–2.0

2.0–20.

0.75

0.25

a. Neutron field characteristics are applicable to certain timeframes as indicated in Table 6-9.

sources), the correction for the ratio of dose equivalent H to personal dose equivalent HP(10)

conversion coefficients of 1.35 in the 2 to 20 MeV range (Table 6-10) should be applied.

TEDs have a lower energy threshold than NTA film (ICRU 2001). Therefore, fewer neutrons will be 

undetected. However, with a threshold of about 100 keV, the underestimate due to energy response 

should be less than 15%, resulting in a bias of 0.87. Additionally a factor of 1.9 between the NTS 

quality factor for neutrons from 0.1 to 2 MeV (10.49), and the currently recommended radiation 

weighting factor, wR, of 20 for the same neutron energy range should be applied as appropriate

(based on energy range). 

On the other hand, albedo dosimeters calibrated with isotopic sources in low-scatter conditions are 

likely to overestimate operational dose if more scatter and moderation could be involved in the 

operational spectra. It is not recommended that credit be taken for potential over-response of albedo 

dosimeters. The bias for albedo dosimeters is to be assigned as 1. Additionally an appropriate ICRP 

correction factor should be assigned (see Table 6-10). This value is recommended for the Panasonic 

UD-809 measurements as well because the detection process is albedo-based. 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 52 of 121

No single individual had access to areas in which there was potential for neutron exposure. Work in 

these areas always involved pairs of workers; two knowledgeable persons had to be involved with 

potential operational activities. This group probably involved only a few hundred workers and was 

limited to persons who worked on specific tasks in specific areas [26]. Neutron doses, for the most 

part, were low. In addition, neutron exposure was not possible without a concomitant gamma 

exposure. Although information on nominal

If workers were unmonitored for fast neutrons, based on NTS personal dosimeter issue practices, and 

if there was no indication of exposure based on the thermal-neutron-sensitive component, it is highly 

unlikely that neutron exposure occurred [34]. If a worker’s duties did not involve access to fissile 

materials or isotopic neutron sources, neutron exposures should not be considered in dose 

reconstruction. Therefore, since 1966, dose reconstructors should not consider missed neutron dose 

for NTS personnel who were not monitored specifically for neutrons. However, dose reconstructors 

must not ignore a positive indication of neutron exposure from an individual’s dose or work records.

neutron-to-gamma ratios appears later in this document, 

in practice these ratios can vary widely. Caution should be exercised before applying these ratios to 

gamma exposures for those who had worked in these operations (Table 6-9). Application to low-level 

photon exposures that were not associated with neutron source operations could lead to unrealistic 

estimates of neutron exposure [33]. 

6.4.3.2 Neutron-to-Photon Dose Ratios

The use of NTA film at NTS began in 1961 (REECo 1961). Prior to that time, an estimation of neutron 

doses to workers known

Table 6-17. Neutron-to-photon dose ratios for isotopic 

sources. 

to have worked with neutron emitting sources or in operations where 

neutrons were present (Table 6-16) could be made from the recorded photon dose, together with an 

appropriate neutron-to-photon dose ratio. The nominal neutron-to-photon dose ratios for isotopic 

neutron sources are listed in Table 6-17. These are based on literature values for the sources free in 

air. However, these values are not appropriate for estimation of the neutron dose from recorded 

photon dose because the photon usually normally includes a large fractional contribution from other 

photon sources that are not accompanied by neutrons. Therefore, if the neutron exposure is 

associated with the use of isotopic sources, use of a neutron-to-photon dose ratio of 5 is 

recommended. These operations include neutron instrument calibration and down-hole well logging.

Neutron sources

Neutron-photon dose 

equivalent rate ratio

Bare Cf-252 20.

Cf-252 moderated by 15 cm D2O 5.6

PuBe-238 29.a

Am-241 >5.b

a. Mayer, Otto, and Golnik (2004). 

b. ISO (2001).

Of the other neutron-related operations listed in Table 6-16, only the nuclear device assembly 

conducted in Area 6 and nuclear explosive assembly using special nuclear material were carried out 

prior to the introduction of personal dosimetry in 1961 (Allen and Schoengold 1995). Specific 

information on the neutron-to-photon dose equivalent ratios is not readily available for these 

operations at NTS. However, similar operations were carried out at the Pantex Plant. Based on 

analysis of dosimetry records at Pantex, as well as at the Hanford and Savannah River Sites, where 

similar materials were handled, resulted in a recommendation in the Pantex TBD on Occupational 

External Dose (ORAUT 2006c) that a neutron-to-photon dose equivalent ratio of 2.5 be adopted for 

Pantex workers.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 53 of 121

Prior to 1961, it is, therefore, recommended that a neutron-to-photon dose equivalent ratio of 2.5 be 

used for NTS workers who were involved in nuclear device assembly or nuclear explosive assembly 

using special nuclear material for neutron dose estimation during the periods when these operations 

were conducted based on the NTS records

Similar guidance on the use of a ratio of 2.5 should apply to low-level TRU waste operations (which 

were conducted beginning in 1970) because they involved special nuclear materials, but only in the 

absence of specific neutron dosimeter results.

. It should, however, be stressed that very few NTS 

employees were actually involved in these operations. These activities were primarily performed by 

national laboratory workers [25]. After 1961, the neutron dosimeter results should be the basis for 

neutron dose estimates. Only when there is no result in the dose record, and it was known that the 

worker was involved in such operations for a particular monitoring period, should the ratio of 2.5 be 

applied to estimate neutron dose.

Reactor operations in Areas 25 and 26 were conducted remotely, so workers were protected by a 

significant separation distance and by shielding in the case of the ramjet operations in Area 26 (Allen 

and Schoengold 1995; DOE 1995b). Moreover, the workers wore neutron dosimeters during reactor 

operation. When the reactor was shut down, the neutron production was drastically reduced to the 

point that the photon dose from fresh fission and activation products far exceeded any remaining 

neutron production. Because of this and the standard use of personal neutron dosimetry, the use of 

neutron-to-photon ratios is not recommended for these reactor operations.

6.5 CLAIM ANALYSIS METHODS

Guidance in external radiation dose reconstruction is presented in ORAUT (2006d). NTS-specific 

recommendations are presented in the following sections. Basics of the dosimetry system through 

different types and operational periods are found in Table 6-1. There were no ongoing processes at 

NTS. The events that occurred in the various operations were discrete occurrences. The site 

description (ORAUT 2004) includes a list of radionuclides of concern for various NTS activities and 

summarizes job descriptions for certain categories of workers that can help the dose reconstructor 

determine the most suitable method to use when evaluating the claim (i.e. minimizing, maximizing, or 

best estimate). 

6.5.1

Beginning in 1987, NTS occupational exposures were recorded in terms of personal dose equivalent, 

HP(d). For exposures during the period from 1951 to 1986, dose reconstructors should use the 

recorded photon dose values in terms of exposure, together with the Exposure to Organ Dose 

coefficients in Appendix B of NIOSH (2006) to determine organ dose. Since 1987, the recorded 

values are in terms of HP(10), and the Deep-Dose-Equivalent-to-Organ-Dose conversion factors 

(NIOSH 2006, Appendix B) should be used. 

Photon Dose

An exception to this guidance is at the NRDS facility during the period from 1970 to 1972 where a 

TLD was used and the Deep-Dose-Equivalent-to-Organ-Dose conversion factors (NIOSH 2006, 

Appendix B) should be used. 

Any recorded doses less than the LOD/2 for the era of the dosimetry should be included as missed 

dose instead of recorded dose (NIOSH 2006). 

Energy Range. For external dose reconstruction, if the conditions of exposure (work area, operation, 

etc.) are unknown, dose reconstructors should use the assumption favorable to claimants that photon 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 54 of 121

energies are between 30 and 250 keV (NIOSH 2006). If the exposure was due to fresh fallout (e.g., 

early reentry teams), it would be reasonable and still favorable to claimants to assume that 75% of the 

photon dose was from photons with energies above 250 keV (Kathren 2004; Coryell and Sugarman 

1951; Nelms and Cooper 1959). If there is adequate documentation linking exposures to a particular 

NTS work area and/or operation, Table 6-9 presents guidance on a reasonable allocation of the 

recorded exposure or personal dose equivalent to the energy groups 30 to 250 keV and greater than 

250 keV.

The multi-element dosimeter was introduced in 1966 (DeMarre 2002). However, from 1961 to 1966, 

when performing a best estimate evaluation of external dose, an additional contribution equivalent to

25% of the total dose should be added in the range of 30 to 250 keV to account for low-energy 

photons attenuated by the lead filter that covered a portion of the film.

Default values are 100% 30 to 250 keV for efficiently processing a claim or 25% 30 to 250 keV and 

75% greater than 250 keV for a reasonable application. Table 6-13 provides more specific values that 

can be used if the location of the worker is clearly identified.

 This is not considered an 

uncertainty and should be assigned to both the measured and the missed dose when applying a best 

estimate methodology (NIOSH 2006).

Bias. The bias should be applied to both the reported and the missed dose, when applying the bias 

increases the dose being assigned (NIOSH 2006). 

Uncertainty. To efficiently process a claim as non-compensable an uncertainty of 1.3 can be applied 

as maximizing to the measured dose (NIOSH 2006). For a reasonable estimate, an uncertainty 

based on the GSD in Table 6-1 could be applied to the measured dose.

Missed Dose

Special Situations

. Missed photon dose can be assigned based on the MDL/2 method and project 

guidance regarding reported dose under the MDL/2, and the number of exchange periods (NIOSH 

2006) given in Table 6-1 for the dosimetry systems or the dose of record provided by DOE. For many 

workers, the number of dosimeter exchanges will be greater than the routine monthly exchange. The 

dose reconstructor should evaluate the exchange cycles and assign the appropriate number of zero 

cycles or default values in Table 6-1 as applicable.

Noble Gas. The potential for noble gas exposure during post-test drilling was primarily limited to 

radiation technicians, drillers, and roughnecks. Such exposures were drastically reduced with the 

introduction of blow-out preventers in 1964 (LRL 1964, 1966). Reentry after tunnel events also had 

the potential for noble gas exposure to reentering miners and other crew members. Ventilation was 

usually introduced into tunnels following the events to reduce the potential for such exposure prior to 

reentry. However, if the ventilation was not working, or had not been provided at all, the potential for 

exposure existed. It should be noted that the number of individuals involved in such operations was 

small [35].

Unmonitored Workers. For potentially exposed unmonitored workers prior to 1958, an estimated dose 

favorable to the claimant can be assigned based on the 50% dose noted in the last column of Table 

6-11. This average annual dose can be adjusted during dose reconstruction to reflect the actual 

annual employment of the exposed person as verified by DOL. Many of the unmonitored workers 

were hired for a specific project and then terminated at the end of the project. Additionally, a missed 

dose of 20 mrem/month (LOD/2) can be assigned as favorable to the claimant. The review of ORAUT 

(2004) should help determine the applicability of assigning 30 mrem per month missed dose to the 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 55 of 121

individual. For example, an administrative worker in Mercury has very limited potential for exposure. 

This should be reflected in any reported dose after April 1957 when universal badging was instituted. 

With all reported dose being zero and no job change, a case can be made that it is reasonable to 

assign the unmonitored dose and no missed dose at 20 mrem/month for an administrative worker in 

Mercury.

6.5.2

If an electron component needs to be assigned based on the cancer location or type, project guidance 

can be found in the current revision to ORAUT (2005), and site-specific guidance can be found in 

Appendices A and B. These sections provide the dose reconstructor with many options dependent on 

the circumstances of the energy employee (e.g., what ratio to apply depends on duties, location, and 

timeframe of employment). For the efficient processing of a claim, the following general guidelines 

can be applied:

Beta Dose

• Before 1966, these ratios can also be applied to recorded photon dose and assigned as 

electron dose unless there is indication in the file that the employee was involved in reentry 

activities after an event. Appendices A and B provide areas and timeframes when a different

ratio might be more appropriate. General guidelines are as follows:

– At less than 3 days after an event, beta dose is 5 times the recorded photon dose.

– At 4 to 42 days after an event, the beta dose is 10 times the recorded photon dose.

– At more than 42 days after an event, the beta dose is 5 times the recorded photon dose.

These provisional dose ratios take no credit for shielding or attenuation of beta particles by 

clothing and assume a 1-m distance from the fission product field with no overburden, 

leaching, vegetation, or other material or weathering action that would reduce the ratio. For 

specific circumstances, other options from ORAUT (2006d) can be justified and applied by the 

dose reconstructor. To efficiently process a claim, a value of 3 to 5 can be justified as 

reasonable, since many of the parameters listed above cannot be quantified for the claim (i.e. 

distance of the worker from the source).

• For a minimizing approach in a compensable claim, the photon dose can also be assigned as 

the electron dose. This can be used when no particulars are available regarding worker 

location and timeframe of exposure (other than exposure prior to 1966). 

• In a few cases prior to 1966, experimental dosimetry can be found in the files provided by 

DOE. These values should be doubled.

• After 1966, for a worker with reported beta or skin dose and no or limited information on duties 

and locations, the dose reconstructor should assign the reported electron dose.

• For a reasonable approach where the location of the employee at a specific point in time can 

be determined, the ratios in Attachment B for contaminated surfaces and immersion in a cloud 

are appropriate. Dependent on the type of contamination environment that is assigned to the 

worker, credit can be assigned for shielding and attenuation, and geometry can be considered, 

when determining the organ dose correction factor.

Because of the uncertainties associated with the application of beta-photon ratios to estimate potential 

beta exposures to NTS workers, dose reconstructors should only use these ratios when other relevant 

information, such as the optical density under the open area of a film badge, is not available. 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 56 of 121

For the beta produced by the decay of fission and activation products, the maximum energy typically 

does not exceed 3 MeV, and the range of a 3-MeV particle in air is approximately 36 ft. Fission 

products with very short half-lives can exceed the maximum energy of 3 MeV, but these can be 

ignored for all practical purposes. Therefore, an individual at a distance greater than 36 ft from a 

fallout field would not receive an external dose from beta radiation associated with the decay of 

radionuclides produced by fission and fission-produced activation products. Similarly, an individual 

exposed to beta particles with energies below 70 keV would receive no beta dose to the skin because 

beta particles with energies below 70 keV have insufficient energy to penetrate the cornified outer 

layer of the skin [36]. 

Bias. The bias should be applied to both the reported and the missed dose, when applying the bias 

increases the dose being assigned. Bias from Table 6-1 should only be applied to electrons during 

those eras where an LOD for electrons is identified.

Uncertainty

Special Situations

. To efficiently process a claim as non-compensable an uncertainty of 2 can be applied as 

maximizing to the measured beta or electron dose. For a reasonable estimate, an uncertainty based 

on the GSD in Table 6-1 could be applied to the measured dose. A minimizing approach is the include 

no uncertainty.

Unmonitored Workers - Prior to universal badging, only workers directly involved with the tests were 

provided with dosimetry (DeMarre 2002). From Table 6-11, photon dose is assigned to unmonitored 

workers based on the 50% dose to those who did receive monitoring. When the workers photon dose 

is assigned in this manner, it is favorable to the claimant to assign an electron dose equal to the 

unmonitored photon dose. This likely contributes a greater electron component that the worker 

received given that the worker status was unmonitored. A minimizing approach is to assign no 

unmonitored electron dose. Given the range of electrons in air and the likelihood of an unmonitored 

worker being within that range, this is a reasonable approach [37]. 

6.5.3

Dose reconstructors should convert recorded neutron dose to HP(10) using the bias values that 

appear in Table 6-1, and use the Deep-Dose-Equivalent-to-Organ-Dose conversion factors from 

Appendix B of NIOSH (2006) to calculate the appropriate organ doses. In addition to the bias, the 

factors in Table 6-10 for conversion from NCRP Report 38 (NCRP 1971) and ICRP Publication 60 

(ICRP 1991) neutron factors are to be applied. Table 6-16 provides neutron field characteristics 

associated with NTS operational areas and Table 6-9 provides the timeframe for application.

Neutron Dose

Energy. Assuming that 100% of the neutron doses were delivered by neutrons in the 0.1 to 2-MeV 

range is favorable to claimants. Table 6-16 includes neutron energy ranges that can be applied if 

more information is available on location and activities of the worker.

Missed Dose

“There was a thermal neutron component in the NTS film badge packet from 1966 through 1986 to 

record neutron dose. Since as every film badge packet issued on site had this component regardless 

of potential for exposure, monitoring data is not necessarily an indicator of exposure to thermal 

. Assign any missed dose based on the number of “OT” (other than thermal) neutrons

found in the records provided by the DOE. The following template language to address thermal 

neutrons has been developed and should be included in any report where the worker was at NTS for 

any period between 1966 and 1986 [38]. 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 57 of 121

neutrons during routine activities. Most of the neutron exposure at NTS in this time frame occurred in 

Area 25 (test reactor area). Low level exposure to neutrons in Area 12 would only intermittently have 

occurred and only involved workers who directly handled the fissile materials used for testing. During 

this activity, fast neutron monitoring was conducted using NTA film and thermal neutrons were 

monitored using the standard NTS film badge. Since Mr. XXX’s responsibilities involved laying 

conduit in preparation for the tests and subsequent recovery of equipment and he was not monitored 

for fast neutrons, it is unlikely that he was involved in the placement of the fissile materials; therefore, 

neutron exposures were not considered for this dose reconstruction.” Starting in 1961 NTA film was 

assigned to those workers with the potential for exposure to neutrons (DeMarre 1993). Prior to that 

based on Table 6-9, it may be appropriate to assign neutrons based on a neutron-to-photon ratio 

applied to the reported photon dose.

Dose reconstructors should ignore neutron exposure unless there is evidence that the claimant was 

within 6 km of one or more of the atmospheric test detonation points at time of detonation.

Bias. The bias should be applied to both the reported and the missed dose, when applying the bias 

increases the dose being assigned. When using the neutron-to-photon ratio, the bias used should be 

from the appropriate photon dosimeter era and the correction factor (conversion from NCRP to ICRP) 

applied based on the information in Table 6-10.

Uncertainty. To efficiently process a claim as non-compensable an uncertainty of 1.3 can be applied 

as maximizing to the measured dose. For a reasonable estimate, an uncertainty based on the GSD in 

Table 6-1 could be applied to the measured dose.

Neutron–Photon Ratio. The workplace neutron fields for specific types of nuclear weapons 

components are classified. Unclassified information on neutron spectra from nuclear weapons 

components is not available, but there are two sources, both with significant components above 2 

MeV. Before about 1960, nuclear weapons could have contained 210PoBe or 238PuBe initiators (DOE 

1997) with a higher neutron dose component relative to the measured photon dose. These initiators 

were not used after 1960. Prior to 1961, it is, therefore, recommended that a neutron-to-photon dose 

equivalent ratio of 2.5 be used for NTS workers who were involved in nuclear device assembly or 

nuclear explosive assembly using special nuclear material for neutron dose estimation during the 

periods when these operations were conducted based on the NTS records

For exposures after 1960, if neutron dose information is not specifically available for those involved 

with final assembly and arming operations, photon exposure records, together with neutron-to-photon 

dose ratios can be used. The neutron-to-photon ratios can be derived from the experience at the 

Pantex Plant, where weapons assembly operations were conducted. Analysis of dose records for 

each Pantex worker with a positive neutron dose greater than 50 mrem for the period from 1993 to 

2003 yields a geometric mean of 0.81 and GSD of 1.51. An upper 95th-percentile value of 1.6 should 

be used for the neutron-to-photon dose ratio (ORAUT 2006c).

. Assembly occurred in 

Area 27, usually using the scientists and technicians from their respective laboratory facilities.

6.6 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction. 

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 58 of 121

[1] DeMarre, Martha E. Nuclear Testing Archive. Manager. October 2003.

Ms. DeMarre, directly and through a large number of background documents, has provided a 

chronicle of the NTS radiation protection program for both external and internal exposure. 

Documents such as the Defense Nuclear Agency Reports (DNA-6000f through DNA-6040f) 

have provided details related to the radiation protection programs for each on the test series.

[2] DeMarre, Martha E. and Griffith, Richard V. Nuclear Testing Archive, and ORAU Team. 

Manager and Principle Consultant. January 2004.

Beta doses are calibrated and measured in terms of absorbed dose. “Beta-particle fields shall 

be calibrated in terms of absorbed dose

[3] Griffith, Richard V. ORAU Team. Principle Consultant. April 2007. 

at a tissue depth of 7 mg/cm2 ...” (DOE 1986a). The 

special unit of absorbed dose is the rad. Dose equivalent, H, is defined as the product of 

absorbed dose and quality factor, Q, or more recently, radiation weighting factor, WR. The 

quality factor and radiation weighting factor for both Q and WR have been set by the 

International Commission on Radiological Protection equal to 1. Therefore, the beta dose 

equivalent and absorbed dose are numerically equal. Therefore, the beta doses can also be 

reported in dose equivalent with the special unit of rem. 

After review of the dosimetry practices that had been used to determine the values to be 

entered in the dosimetry records, the bias factor represents the under- or overestimate of the 

measurement quantity by the recorded dose. By the convention adopted by the NTS dose 

reconstruction project, a bias factor less than 1.0 is used when the recorded value is believed 

to be an underestimate of the measurement quantity, while a bias factor greater than 1.0 is 

used when the recorded value is an overestimate of the measurement quantity. Simply put, 

bias factor = recorded value/measurement quantity.

[4] Griffith, Richard V. ORAU Team. Principle Consultant. April 2007. 

NIOSH (2006) specifies that the value of the measurement quantity should then be multiplied 

by the dose conversion factors in Appendix B to obtain the organ dose.

[5] DeMarre, Martha E. Nuclear Testing Archive. Manager. October 2003.

The dosimetry processing procedures, including the use of control films, has been outlined in 

REECo Environmental Sciences Standard Procedures for the years 1952 through 1985. 

These procedures have been compiled by Ms. DeMarre and are available. However, because 

of the number of procedures involved, they are not referenced individually.

[6] DeMarre, Martha E. Nuclear Testing Archive. Manager. October 2003.

This is a matter of NTS record.

[7] DeMarre, Martha E. and Griffith, Richard V. Nuclear Testing Archive, and ORAU Team. 

Manager and Principle Consultant. January 2004.

The temporary assignment of major laboratory and contractor employees at the NTS for 

periods of up to several weeks, perhaps punctuated by a return home over a weekend or for a 

few days, was a common practice and well known in the AEC/DOE contractor community.

[8] DeMarre, Martha E. Manager, Nuclear Testing Archive. January 2004.

Ms. DeMarre has substantiated the practice of contractor employees exchanging NTS issued 

dosimeters for contractor badges on completion of their assignments.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 59 of 121

[9] Griffith, Richard V. ORAU Team. Principle Consultant. February 2007.

This is based on the experience of the data entry technicians during data entry. According to 

Cheryl Moore, Dade Moeller and Associates, NIOSH Project Data Entry Supervisor, often the 

LLNL or LANL records identify the NTS results as "NTS" or "visitor".

[10] DeMarre, Martha E. Manager, Nuclear Testing Archive. February 2006.

TTR is the Tonopah Test Range. Sandia National Laboratory handled the dosimetry (if there 

was any) for the Tonopah Test Range. REECo had a Human Resources Office outside the 

NTS and one in town. There was a REECo medical clinic in Mercury and one in Las Vegas 

(Leased space until 1994).

When a REECo employee "processed in," the employee would usually start the process 

downtown. Downtown employees completed the process downtown. If the employee worked 

at NTS or Tonopah, they would usually report to the REECo Human Resources Office outside 

the gate. When they were ready to go to medical (which was inside at Mercury), they would 

have to get a dosimeter at Building 1000 (guard building - outside the gate). Once they had 

their badge, they would go to REECo medical (Mercury Building 650) and have their physical. 

They would physically be on the NTS for a very short period of time (hours). 

They are actually listed as NTS visitors, where there is a place for a visitor indicator in the 

record. They do not have a permanent NTS security badge for TTR employment.

TTR employment was considered "Work for Others" (WFO) (in support of Sandia and the 

U.S. Air Force).

[11] Griffith, Richard V. ORAU Team. Principle Consultant. February 2007.

A hot particle is essentially a point source of radiation. If that particle is not deposited directly 

on the dosimeter, the radiation field from the particle spreads so that the dosimeter cannot 

distinguish between the radiation field from the particle and that from a distributed source. 

Therefore, a hot particle that was deposited on the body at even a small distance from the 

dosimeter could not be identified.

[12] Griffith, Richard V. ORAU Team. Principle Consultant. February 2007.

Hot particles will be removed in a short time (usually less than a day) after deposition on the 

body or clothing by washing or removal of the clothing. Once the particle has been removed, 

there is no way of knowing that it had been deposited unless associated radiation damage to 

the skin and underlying tissue manifested itself as an area of physical damage or a lesion.

[13] Griffith, Richard V. ORAU Team. Principle Consultant. February 2007.

During interviews with people who were involved in NTS operations, including Jay Brady, there 

were a number of references to occasions when workers had intentionally left dosimeters in 

locations where they knew there would be little or no dose. One of the reasons commonly 

given was concern about reduction in pay that might be associated with work restrictions 

imposed if dose limits were exceeded. It should be noted that reports of this practice were not 

limited to NTS activities.

[14] DeMarre, Martha E. and Griffith, Richard V. Nuclear Testing Archive, and ORAU Team. 

Manager and Principle Consultant. February, 2004.

The development of dosimetry technology, not only at the NTS but elsewhere in the AEC/DOE 

community, is documented in a myriad of reports by REECo, the national laboratories, the 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 60 of 121

military, and other contractors. This development is reflected in several of the reports in the 

reference list, and is not necessary here.

[15] Kathren, Ronald L. ORAU Team. Consultant. January 2004.

The potential large uncertainty (a factor of 2) is based on professional experience. The 

recommendation that “dose reconstructors should double the reported value to ensure 

favorability to claimants and to account for uncertainties” follows from this experience.

[16] Griffith, Richard V. ORAU Team. Principle Consultant. February 2007.

The sensitive element of both film and TLD dosimeters used to detect incident beta particles is 

covered with a thin layer of protective material (paper, plastic, etc.). The range of a 0.015-

MeV beta particle in unit density material is 4 × 10-3 mm, so the beta particles with energies 

≤0.015 MeV would be absorbed in the inert protective material and not be detected.

[17] DeMarre, Martha E. and Griffith, Richard V. Nuclear Testing Archive, and ORAU Team. 

Manager and Principle Consultant. February 2004.

Mechanisms for production of neutrons are limited and well defined:

• Nuclear fission

• Alpha,neutron reactions with a small number of light nuclides (lithium, boron, beryllium, 

etc.)

• Accelerator production

The facilities or operations at the NTS that are capable of producing neutrons have been 

identified in Section 2 (ORAUT 2004c) and are summarized briefly in this section. Because of 

the nature of these operations, and the associated hazards, access of individuals has been 

carefully controlled.

NTA film, albedo TLDs, and track etch dosimeters were not issued to all personnel at the NTS. 

They were issued only to individuals that might have a potential for exposure to neutrons.

The brief reactor runs had a very small potential for neutron exposure. The higher potential 

was from sources (such as Pu-Be & 252Cf) to an extremely small number of people. The tasks 

included instrument calibration and well logging. The number of workers with a potential for 

exposure would be extremely low.

[18] Griffith, Richard V. ORAU Team. Principle Consultant. February 2004.

The use of field calibrations is and has been standard practice in radiation protection for many 

years. It allows direct comparison of the response of dosimeters and reference or survey 

instruments in the occupational environment and eliminates, among other things, uncertainties 

about the effects of energy response because the operational field becomes the calibration 

field as well.

[19] Griffith, Richard V. ORAU Team. Principle Consultant. February 2004.

Neutron calibration can be performed with isotopic sources, reactor beams, or accelerators. 

However, use of facilities such as reactors and accelerators for calibration of dosimeters and 

survey instruments involves a significant commitment of resources and labor. They are 

normally only used for special calibration requirements such as establishing the energy 

response of instruments and dosimeter designs. Once that has been done, simpler, less 

labor-intensive methods are used to confirm proper operation of neutron detection equipment. 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 61 of 121

Isotopic sources such as 252Cf, 238Pu-Be, etc. are much more readily available to the normal 

radiation protection organization, and are used much more widely.

[20] DeMarre, Martha E. and Griffith, Richard V. Nuclear Testing Archive, and ORAU Team. 

Manager and Principle Consultant. March 2004.

Operational field calibrations for neutrons require levels of a few tenths of a millirem so the 

measurements can be conducted in a reasonable time. From discussions with NTS radiation 

protection specialists such as Jay Brady and Joe Wells, or in going through the NTS records, 

there is no indication that such field calibrations were ever attempted.

[21] Griffith, Richard V. ORAU Team. Principle Consultant. March 2007.

Because of variations in track etch detector materials, etc., the practical detection level for fast 

neutrons (from 252Cf, 238Pu-Be, etc.) is higher than that achieved under highly controlled 

conditions – typically about 25 mrem. However, for moderated spectra, a significant fraction of 

the neutrons will have energies below the energy detection threshold for the TEDs. In 

addition, because the factor to be applied to NTS neutron dose in the 0.1- to 2-MeV region is 2 

as a result of the difference between NCRP and ICRP dose conversion factors (Table 6-10), 

the recommended MDL is 50 mrem.

[22] DeMarre, Martha E. and Griffith, Richard V. Nuclear Testing Archive, and ORAU Team. 

Manager and Principle Consultant. March 2004.

Because of the scale and design of operations, and the time to reentry, exposure to 

radioiodines and noble gases during tunnel reentries was limited to those actually involved in 

the reentry – miners and radiation technicians.

[23] Griffith, Richard V. ORAU Team. Principle Consultant. March 2007.

Fission products are formed directly so that there are at least two fission product atoms formed 

from each fission. Activation products are formed as a secondary process through interaction 

of the fission product neutrons with adjacent materials (soil, structures, etc.). Activation 

product formation is geometry dependent so that fewer activation product atoms are formed 

from an air burst than a surface burst. Moreover, the activation products are distributed with 

depth in soil because they originate primarily by interactions with neutrons that penetrate into 

the soil rather than in deposition from the atmosphere. Thus, beta doses from activation 

products are reduced because most of the activation products are deeper in the soil than the 

range of the emitted beta particles.

[24] Griffith, Richard V. ORAU Team. Principle Consultant. March 2007.

Attachment D illustrates, using two different methods, that personnel doses at distances 

greater than 6 km would be less than 1 mrem.

[25] DeMarre, Martha E. Nuclear Testing Archive. Manager. April 2006.

The device was assembled by the specialists from the laboratories (LANL, LLNL, etc.).

[26] DeMarre, Martha E. and Griffith, Richard V. Nuclear Testing Archive, and ORAU Team. 

Manager and Principle Consultant. February 2004.

Neutron sources were used for very specific jobs such as instrument calibration and well 

logging. These jobs required specialized training, including the potential hazards associated 

with use of the sources, and involved only a few individuals such as radiation technicians. 

When they were in use, individuals who were not directly involved in the task were kept away 

from the sources and at a safe distance.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 62 of 121

[27] DeMarre, Martha E. Nuclear Testing Archive. Manager. October 2003.

Ms. DeMarre has provided information about personnel dosimeter issuance. 

[28] Griffith, Richard V. ORAU Team. Principle Consultant. March 2007.

The error introduced in using a 1-to-1 rad-to-roentgen relationship and ignoring the conversion 

of 100 rad = 114 R is 14 %. However, the uncertainties caused by the significant beta 

geometry and energy dependence of the beta-photon ratios are much larger and can exceed a 

factor of 2 or more. Therefore, 14% becomes acceptable in the context of other much larger 

errors.

[29] Griffith, Richard V. ORAU Team. Principle Consultant. March 2007.

Immersion in a cloud was generally associated with radioactive gases, which are normally 

dispersed quite rapidly in an open environment. Under normal conditions, this would not be 

common. Most exposures would have resulted from surfaces (ground, etc.) that were 

contaminated by postshot fallout. The exception would be exposure following a release of 

gases during postshot drilling.

[30] Griffith, Richard V. ORAU Team. Principle Consultant. March 2007.

Using data from Keith Eckerman and the EPA Federal Guidance Report-13 code (Eckerman 

et al. 1999) it is possible to calculate beta-photon ratios for the fallout inventories generated by 

Hicks (1981a to 1981i). These ratios can then be applied, with care, to the photon doses 

obtained from the personal dosimetry records.

[31] Griffith, Richard V. ORAU Team. Principle Consultant. March 2007.

Beta-photon ratios are highly dependent on geometry, including any intervening material (air, 

etc.), beta and photon energies, and other factors. Although calculations can be made for 

well-defined situations, application of the results can result in significant errors if the actual 

exposure conditions deviate significantly from the assumptions that were used for the 

calculations.

[32] Griffith, Richard V. ORAU Team. Principle Consultant. March 2007.

The energy spectrum of fast neutrons can easily be degraded by intervening materials and 

loss of energy through scattering from materials in the vicinity of the source. This occurs, for 

example, when there is a significant amount of concrete, soil, etc. near the source of neutrons.

[33] Griffith, Richard V. ORAU Team. Principle Consultant. March 2007.

Application of neutrons gamma ratios to the recorded photon dose from the individual 

monitoring records assumes that the total photon dose was received from a given mixed 

photon-neutron source environment. However, if a significant portion of the photon dose for 

the monitoring period was received in an environment that did not have associated neutrons 

(e.g., a neutron-free field with only photon and beta contamination), the estimated neutron 

dose would be higher than actually received.

[34] DeMarre, Martha E. and Griffith, Richard V. Nuclear Testing Archive, and ORAU Team. 

Manager and Principle Consultant. March 2004.

Personnel were issued fast neutron dosimeters if they were known to have been working with 

or had access to neutron sources. Moreover, all personnel dosimeters after 1961 had some 

thermal neutron sensitive component. If (1) an individual was not issued a fast neutron 

dosimeter and (2) there was no indication of a thermal neutron exposure on the thermal

neutron sensitive component, it was highly unlikely that there had been a neutron exposure. 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 63 of 121

Such exposure would have been reflected by an indication on the thermal dosimeter 

component.

[35] DeMarre, Martha E. and Griffith, Richard V. Nuclear Testing Archive, and ORAU Team. 

Manager and Principle Consultant. March 2004.

The personnel with potential for noble gas exposure during a postshot drilling operation were 

those whose job assignments called for them to be directly involved in the drilling operation – 

drilling crew members and radiological technicians. Because of the training and skills that

were required for these jobs, the number of people who were involved was small.

[36] Griffith, Richard V. ORAU Team. Principle Consultant. March 2007.

The range of a 70-keV beta particle is about 7 mg/cm2 in tissue, and it cannot penetrate the 

dead or cornified outer layer of the skin. Therefore, there is no beta dose to the sensitive layer 

from beta particles with energies below 70 keV.

[37] Griffith, Richard V. ORAU Team. Principle Consultant. March 2007.

Unmonitored workers are highly unlikely to have been in areas where significant beta fields 

would have existed. More over, in view of the short range of beta particles in air (generally 

less than 100 cm), (1) workers would have had to be quite close to these sources and (2) the 

sources would have been much more intense than contamination background fields. 

Therefore, it is quite reasonable to assign no unmonitored electron dose for unmonitored 

workers.

[38] Arana, Joel. ORAU Team. Senior Health Physicist. March 2007.

NIOSH requested that we elaborate a bit better on our explanation of the thermal neutron 

component in the NTS film badge. This is some proposed wording from Tim Taulbee. It also 

should be noted that the thermal neutron component really isn’t for accident monitoring, so it is 

inappropriate to say so.

[39] Griffith, Richard V. ORAU Team. Principle Consultant. April 2007.

Allen and Schoengold (1995) states, “The Panasonic UD-802 TLDs have well-known and 

documented energy responses.” As a large commercial supplier of dosimeters, this is true of 

all of the Panasonic TLDs, including the UD-809, which is not specifically addressed in the 

REECo document.

[40] Griffith, Richard V. ORAU Team. Principle Consultant. April 2007.

This information is provided in detail in the Characteristics of the Panasonic UD-8xx series 

Thermoluminescent Dosimeters fact sheet from the Panasonic Web site at

http://www.panasonic.com/industrial/other/pdf/dosimeter_types.pdf.

[41] Griffith, Richard V. ORAU Team. Principle Consultant. April 2007.

The neutron response of NTA film at 4 MeV is approximately 2.1 times that at 1 MeV (IAEA 

1990). This means that 4-MeV neutrons, when corrected for a personal dose equivalent 

difference of less that 10%, will create twice as many tracks as 1 MeV neutrons per unit 

HP(10). As a result, if a dosimeter has been calibrated with an alpha,neutron source such as 

plutonium or Am-Be and that dosimeter is used in a fission neutron environment, the number 

of track density is interpreted as a dose that is half the actual fission neutron dose. This 

underestimate is even worse if the field has a significant number of neutron below the 0.5-MeV 

NTA threshold.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 64 of 121

REFERENCES

AEC (U.S. Atomic Energy Commission), 1961, Radiological Safety, Chapter 0524, NTSO-0524-1, Las 

Vegas, Nevada, June 26. [SRDB Ref ID: 12468]

AECL (Atomic Energy of Canada Limited), 1982, Tables of Beta-Ray Dose Distribution in Water, Air 

and Other Media, AECL-7617, September.

Allen, T. E., and C. R. Schoengold, 1995, External Dosimetry Technical Basis Manual for the Nevada 

Test Site and Other Designated Locations, (548)-6.4.1, Reynolds Electrical and Engineering 

Company, Las Vegas, Nevada, December. [SRDB Ref ID: 14099]

Barss, N. M., 2000, Methods and Applications for Dose Assessment of Beta Particle Radiation, SAIC001/2024, Science Applications International Corporation, January. [SRDB Ref ID: 26379]

Barss, N. M., and R. L. Weitz, 2006, “Reconstruction of External Dose From Beta Radiation Sources 

of Nuclear Weapon Origin,” Health Physics, volume 91, number 4, pp. 379–389. [SRDB Ref 

ID: 26776]

Bechtel Nevada, 2001, Dosimetry Issue/Return, Organization Instruction, OI-0441.024, 31 October, 

2001.

Beck, H. L., 1980, Exposure Rate Conversion Factors for Radionuclides Deposited on the Ground, 

EML-387, U.S. Department of Energy, Environmental Measurements Laboratory, New York, 

New York, July.

Becker, K., 1966, Photographic Film Dosimetry, The Focal Press, London, England.

Black, R. H., 1962, “Some Factors Influencing the β-Dosage to Troops,” Health Physics, volume 8, pp. 

131–141.

Boone, F. W., E. D. Bennett, and S. B. Adams, 1970, Progress Report, NRDS Personnel Dosimetry 

Program, Statistical Comparison of TLD and Film Dosimeter Data, 33-37, Pan American World 

Airways, Nuclear Rocket Development Station Project, Jackass Flats, Nevada, July 24. 

[SRDB Ref ID: 13988]

Brady, W. J., and G. K. Iverson, 1968, Combination Personnel Dosimeter and Security Credential 

Holder, NVO-162-27 (revised) , Reynolds Electrical and Engineering Company, Nevada Test 

Site, Mercury, Nevada, September. [SRDB Ref ID: 1477]

Brady, W. J., and A. G. Nelson, 1985, Radiac Instruments and Film Badges Used at Atmospheric 

Nuclear Tests, DNA-TR-84-338, Jaycor, Alexandria, Virginia, September 26. [SRDB Ref ID: 

1510]

Broido, A., and J. D. Teresi, 1961, “Analysis of the Hazards Associated with Radioactive Fallout 

Material – I. Estimation of γ- and β-Doses,” Health Physics, volume 5, pp. 63–69.

Cardis, E., M. Vrijheid, M. Blettner, E. Gilbert, M. Hakama, C. Hill, G. Howe, J. Kaldor, C. R. 

Muirhead, M. Schubauer-Berigan, T. Yoshimura, F. Bermann, G. Cowper, J. Fix, C. Hacker, B. 

Heinmiller, M. Marshall, I. Thierry-Chef, D. Utterback, Y-O. Ahn, E. Amoros, P. Ashmore, A. 

Auvinen, J-M. Bae, J. B. Solano, A. Biau, E. Combalot, P. Deboodt, A. Diez Sacristan, M. 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 65 of 121

Eklof, H. Engels, G. Engholm, G. Gulis, R. Habib, K. Holan, H. Hyvonen, A. Kerekes, J. 

Kurtinaitis, H. Malker, M. Martuzzi, A. Mastauskas, A. Monnet, M. Moser, M. S. Pearce, D. B. 

Richardson, F. Rodriguez-Artalejo, A. Rogel, H. Tardy, M. Telle-Lamberton, I. Turai, M. Usel, 

and K. Veress, 2005, "Risk of Cancer After Low Doses of Ionising Radiation: Retrospective 

Cohort Study in 15 Countries," British Medical Journal, volume 331, p. 77.

Cheka, J. S., 1954, “Recent Developments in Film Monitoring of Fast Neutrons,” Nucleonics, Volume 

12, number 6, pp. 40–43.

Collison, T. D., 1953, Radiological Safety Operation, WT-702(REF), Field Command, Armed Forces 

Special Weapons Project, Albuquerque, New Mexico, June.

Collison, T. D., 1955, Radiological Safety, Field Command, Armed Forces Special Weapons Project, 

Albuquerque, New Mexico, May.

Coryell, C. D., and N. Sugarman, editors, 1951, The Fission Products, National Nuclear Energy 

Series, Book I, Division IV, Part IV, McGraw-Hill, New York, New York.

Cross, W. G., H. Ing, N. O. Freedman, and J. Mainville, 1982, Tables of Beta-ray Dose Distributions 

in Water, Air and Other Media, AECL-7617, Atomic Energy of Canada, Chalk River, Ontario.

DeMarre, M., 1993, 1951-1973-NTS Instrument Summary, Private Communication, U.S. Department 

of Energy, Nevada Operations Office, Las Vegas, Nevada.

DeMarre, M., 2002, "Monitoring of Personnel Exposures in Association with the Nuclear Testing 

Program," Nevada Test Site External Dosimetry Fact Sheet. 

DeMarre, M., 2003, “RE: Information Request,” e-mail to L. Arent (Oak Ridge Associated Universities 

Team), U.S. Department of Energy, Nevada Operations Office, Las Vegas, Nevada, 

September 17. [SRDB Ref ID: 14053]

DeMarre, M., 2006a, "Use of PIC Dosimeter Results at the NTS," private communication with R. 

Griffith (Oak Ridge Associated Universities Team), U.S. Department of Energy, Nevada 

Operations Office, Las Vegas, Nevada.

DeMarre, M., 2006b, "NTPR Data Definitions," private communication with R. Griffith (Oak Ridge 

Associated Universities Team), U.S. Department of Energy, Nevada Operations Office, Las 

Vegas, Nevada.

DeMarre, M., 2006c, "Extremity Monitoring for NTS Workers," private communication with R. Griffith 

(Oak Ridge Associated Universities Team), U.S. Department of Energy, Nevada Operations 

Office, Las Vegas, Nevada.

DeMarre, M., 2006d, "Neutron Exposure Potential for NTS Workers," private communication with J. 

Fix (Oak Ridge Associated Universities Team), U.S. Department of Energy, Nevada 

Operations Office, Las Vegas, Nevada.

DeMarre, M., 2006e, "Exposure Matrix for NTS and PPG Associated Persons Based on Query of 

Historical Dosimetry (1945-1983) Database," e-mail to J. Fix (Oak Ridge Associated 

Universities Team), U.S. Department of Energy, Nevada Operations Office, August 3. [SRDB 

Ref ID: 27080]
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 66 of 121

DNA (Defense Nuclear Agency), 1985, Neutron Exposure for DoD Nuclear Test Personnel, DNA-TR84-405, Science Applications International Corporation, McLean, Virginia, August 15.

DOE (U. S. Department of Energy), 1986a, Department of Energy Standard for the Performance 

Testing of Personnel Dosimetry Systems, DOE Laboratory Accreditation Program for 

Personnel Dosimetry Systems, DOE/EH-0027, Assistant Secretary for Environment, Safety, 

and Health, Washington, D.C., December. [SRDB Ref ID: 12294]

DOE (U.S. Department of Energy), 1982, Radiological Safety, Chapter 0524, NTS-SOP-0524, 

Nevada Operations Office, Las Vegas, Nevada, July 23. [SRDB Ref ID: 13991]

DOE (U.S. Department of Energy), 1986b, Handbook for the Department of Energy Laboratory 

Accreditation Program for Personnel Dosimetry Systems, DOE/EH-0026, Assistant Secretary 

for Environment, Safety, and Health, Washington, D.C., December. [SRDB Ref ID: 12293]

DOE (U.S. Department of Energy), 1988, Radiation Protection for Occupational Workers, Order 

5480.11, Office of Environment, Safety and Health, Washington D.C., December 21. [SRDB 

Ref ID: 8073]

DOE (U.S. Department of Energy), 1994, U.S. Department of Energy Radiological Control Manual, 

DOE/EH-0256T, Rev. 01, Office of Assistant Secretary for Environment, Safety and Health, 

Washington, D.C., April. [SRDB Ref ID: 14856]

DOE (U.S. Department of Energy), 1995a, Department of Energy Laboratory Accreditation Program 

for Personnel Dosimetry Systems, DOE-STD-1095-95, U.S. Department of Energy, 

Washington, D.C., December, 1995.

DOE (U.S. Department of Energy), 1995b, Radiological Effluents Released from Nuclear Rocket and 

Ramjet Engine Tests at the Nevada Test Site: Fact Book, DOE/NV 401, June.

DOE (U.S. Department of Energy), 1997, Linking Legacies: Connecting the Cold War Nuclear 

Weapons Production Processes to Their Environmental Consequences, DOE/EM-0319, Office 

of Environmental Management, Washington, D.C., January. [SRDB Ref ID: 11930]

DOE (U.S. Department of Energy), 2000, United States Nuclear Tests, July 1945 through September 

1992, DOE/NV--209-Rev. 15, Nevada Operations Office, Las Vegas, Nevada, December. 

[SRDB Ref ID: 2433]

DOE (U.S. Department of Energy), 2002, Nuclear Rocket Development Station tested nuclear rocket 

for Mars mission, DOE/NV Information Sheet, Nevada Operations Office, Las Vegas, Nevada.

Dudley, R. A., 1966, “Dosimetry with Photographic Emulsions,” Radiation Dosimetry, Volume 2, F. H. 

Attix and W. C. Roesch, editors, Academic Press, New York, New York.

Durham, J. S., 1992, VARSKIN Mod 2 and SADDE Mod 2: Computer Codes for Assessing Skin Dose 

from Skin Contamination, NUREG/CR-5873, U.S. Nuclear Regulatory Commission, 

Washington, D.C.

Eckerman, K. F., 2006a private communication with R. Griffith (Oak Ridge Associated Universities), 

Oak Ridge National Laboratory, Oak Ridge, Tennessee.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 67 of 121

Eckerman, K. F., 2006b, private communication with R. Griffith (Oak Ridge Associated Universities), 

Oak Ridge National Laboratory, Oak Ridge, Tennessee.

Eckerman, K. F., R. W. Leggett, C. B. Nelson, J. S. Puskin, and A. C. B. Richardson, 1999, Cancer 

Risk Coefficients for Environmental Exposure to Radionuclides, Radionuclide-Specific Lifetime 

Radiogenic Cancer Risk Coefficients for the U.S. Population, Based on Age-Dependent 

Intake, Dosimetry, and Risk Models, Federal Guidance Report No. 13, EPA 402-R-99-001, 

U.S. Environmental Protection Agency, Office of Radiation and Indoor Air, Washington, D.C., 

September. [SRDB Ref ID: 13719]

Ehrlich, M., 1954, Photographic Dosimetry of X and Gamma Rays, Handbook 57, U.S. Department of 

Commerce, National Bureau of Standards, Washington, D.C., August 20. [SRDB Ref ID: 

11309]

Ehrlich, M., 1962, The Use of Film Badges for Personnel Monitoring, Safety Series No. 8, STI-PUB43, International Atomic Energy Agency, Vienna, Austria, June 28.

Firestone, R. B., and L. P. Ekström, 2004, LBNL Isotopes Project - LUNDS Universitet, WWW Table 

of Radioactive Isotopes, Lawrence Berkeley National Laboratory and Lund University, Sweden

http://ie.lbl.gov/toi/.

Fix, J. J., E. S. Gilbert, and W. V. Baumgartner, 1994, An Assessment of Bias and Uncertainty in 

Recorded Dose from External Sources of Radiation for Workers at the Hanford Site, PNL10066, Battelle Memorial Institute, Pacific Northwest Laboratory, Richland, Washington, 

August. [SRDB Ref ID: 305]

Fix, J. J., L. Salmon, G. Cowper, and E. Cardis, 1997, "A Retrospective Evaluation of the Dosimetry 

Employed in an International Combined Epidemiologic Study," Radiation Protection Dosimetry, 

volume 74, pp. 39–53. [SRDB Ref ID: 7769]

Fix, J. J., R. H. Wilson, and W. V. Baumgartner, 1997, Retrospective Assessment of Personnel 

Neutron Dosimetry for Workers at the Hanford Site, PNNL-11196, Battelle Memorial Institute, 

Pacific Northwest National Laboratory, Richland, Washington, February. [SRDB Ref ID: 

5275]

Glasstone, S., and P. J. Dolan, editors, 1977, The Effects of Nuclear Weapons, Third Edition, U.S. 

Department of Defense and U.S. Department of Energy, Washington, D.C.

Griffith, R. V., 2004, Resolution of Question Regarding Quantities Used for Reporting the 

Measurement of Personnel Exposures to Photons at the NTS, 1962 to 1987, Woodbridge, 

California, August 4. [SRDB Ref ID: 13989]

Griffith, R. V., 2005, Development of Work Area and Operation-Dependent Energy Group Photon 

Fractions, Oak Ridge Associated Universities, Oak Ridge, Tennessee.

Griffith, R. V., 2006, "Correction Factors for External Environmental Dose," Private Communication, 

Oak Ridge Associated Universities, Oak Ridge, Tennessee.

Griffith, R.V. and Tommasino, L. 1990, "Etch Track Detectors in Radiation Dosimetry," The Dosimetry 

of Ionizing Radiation, Volume III, Academic Press, pp. 323–426.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 68 of 121

Hadlock, D. E., L. W. Brackenbush, R. V. Griffith, D. E. Hankins, and M. A. Parkhurst, 1988, 

Personnel Neutron Dose Assessment Upgrade, Volume I, Personnel Neutron Dosimetry 

Assessment, PNL-6620 Vol. 1, Battelle Memorial Institute, Pacific Northwest Laboratory, 

Richland, Washington, July. [SRDB Ref ID: 14858]

Hankins, D. E., 1977, “Energy Dependence of Remmeters and Albedo Neutron Dosimeters at 

Energies of Thermal and between 2 keV and 5.67 MeV,” Proceedings of the Fourth 

International Congress of the International Radiation Protection Association, volume 2, pp. 

553–556. [SRDB Ref ID: 13994]

Hankins, D. E., S. G. Homann, and J. Westermark, 1987, Personnel Neutron Dosimetry Using 

Electrochemically Etched CR-39 Foils, UCRL-53833, University of California, Lawrence 

Livermore National Laboratory, Livermore, California. [SRDB Ref ID: 13995]

Hicks, H. G., 1981a, Calculation of the Concentration of any Radionuclide Deposited on the Ground 

by Release from Underground Nuclear Detonations, Tests of Nuclear Rockets, and Tests of 

Nuclear Ramjet Engines, UCRL-53228, University of California, Lawrence Livermore National 

Laboratory, Livermore, California, November. [SRDB Ref ID: 1828]

Hicks, H. G., 1981b, Results of Calculations of External Gamma Radiation Exposure Rates from 

Fallout and the Related Radionuclide Compositions, Operation Ranger, 1951, UCRL-53152-

Pt.1, University of California, Lawrence Livermore National Laboratory, Livermore, California, 

July. [SRDB Ref ID: 1854]

Hicks, H. G., 1981c, Results of Calculations of External Gamma Radiation Exposure Rates from 

Fallout and the Related Radionuclide Compositions, Operation Buster Jangle, 1951, UCRL53152-Pt.2, University of California, Lawrence Livermore National Laboratory, Livermore, 

California. [SRDB Ref ID: 1831]

Hicks, H. G., 1981d, Results of Calculations of External Gamma Radiation Exposure Rates from 

Fallout and the Related Radionuclide Compositions, Operation Tumbler-Snapper, 1952, 

UCRL-53152-Pt.3, University of California, Lawrence Livermore National Laboratory, 

Livermore, California, July. [SRDB Ref ID: 1832]

Hicks, H. G., 1981e, Results of Calculations of External Gamma Radiation Exposure Rates from 

Fallout and the Related Radionuclide Compositions, Operation Upshot-Knothole, 1953, UCRL53152-Pt.4, University of California, Lawrence Livermore National Laboratory, Livermore, 

California, July. [SRDB Ref ID: 1835]

Hicks, H. G., 1981f, Results of Calculations of External Gamma Radiation Exposure Rates from 

Fallout and the Related Radionuclide Compositions, Operation Teapot, 1955, UCRL-53152-

Pt.5, University of California, Lawrence Livermore National Laboratory, Livermore, California, 

July. [SRDB Ref ID: 1838]

Hicks, H. G., 1981g, Results of Calculations of External Gamma Radiation Exposure Rates from 

Fallout and the Related Radionuclide Compositions, Operation Plumbob, 1957, UCRL-53152-

Pt.6, University of California, Lawrence Livermore National Laboratory, Livermore, California, 

July. [SRDB Ref ID: 1839]

Hicks, H. G., 1981h, Results of Calculations of External Gamma Radiation Exposure Rates from 

Fallout and the Related Radionuclide Compositions, Operation Hardtack II, UCRL-53152-Pt.7, 
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 69 of 121

University of California, Lawrence Livermore National Laboratory, Livermore, California, July. 

[SRDB Ref ID: 1846]

Hicks, H. G., 1981i, Results of Calculations of External Gamma Radiation Exposure Rates from 

Fallout and the Related Radionuclide Compositions, Operation Nougat through Bowline, 

1962–1968, UCRL-53152-Pt.8, University of California, Lawrence Livermore National 

Laboratory, Livermore, California, July. [SRDB Ref ID: 1851]

Hicks, H. G., 1982, “Calculation of the Concentration of any Radionuclide Deposited on the Ground by 

Offsite Fallout from a Nuclear Detonation”, Health Physics, volume 42, number 5, pp. 585–

600.

Hicks, H. G., 1984, Results of Calculations of External Gamma Radiation Exposure Rates from Local 

Fallout and the Related Radionuclide Compositions of Selected U.S. Pacific Events, UCRL53505, University of California, Lawrence Livermore National Laboratory, Livermore, 

California, February. [SRDB Ref ID: 1856]

IAEA (International Atomic Energy Agency), 1979, Radiological Safety Aspects of the Operation of 

Electron Linear Accelerators, Technical Reports Series No. 188, Vienna, Austria, March.

IAEA (International Atomic Energy Agency), 1990, Compendium of Neutron Spectra and Detector 

Responses for Radiation Protection Purposes, Technical Reports Series No. 318, Vienna, 

Austria, July 26.

IAEA (International Atomic Energy Agency), 2001, Compendium of Neutron Spectra and Detector 

Responses for Radiation Protection Purposes, Supplement to Technical Reports Series No. 

318, Technical Reports Series 403, Vienna, Austria.

ICRP (International Commission on Radiological Protection), 1987, Data for Use in Protection Against 

External Radiation, ICRP Publication 51, Pergamon Press.

ICRP (International Commission on Radiological Protection), 1991, 1990 Recommendations of the 

International Commission on Radiological Protection, Publication 60, Pergamon Press, Oxford, 

England.

ICRU (International Commission on Radiation Units and Measurements), 1962, Radiation Quantities 

and Units, Report 10a, Bethesda, Maryland.

ICRU (International Commission on Radiation Units and Measurements), 1988, Determination of 

Dose Equivalents from External Radiation Sources–Part 2, Report 43, Bethesda, Maryland, 

December 15.

ICRU (International Commission on Radiation Units and Measurements), 1992, Measurement of 

Dose Equivalents from External Photon and Electron Radiations, Report 47, Bethesda, 

Maryland.

ICRU (International Commission on Radiation Units and Measurements), 1993, Quantities and Units 

in Radiation Protection Dosimetry, Report 51, Bethesda, Maryland.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 70 of 121

ICRU (International Commission on Radiation Units and Measurements), 1998, Conversion 

Coefficients for Use in Radiological Protection Against External Radiation, Report 57, 

Bethesda, Maryland.

ICRU (International Commission on Radiation Units and Measurements), 2001, Determination of 

Operational Dose Equivalent Quantities for Neutrons, Report 66, Bethesda, Maryland.

Ing, H., and S. Makra, 1978, Compendium of Neutron Spectra in Criticality Accident Dosimetry, 

Technical Reports Series No. 180, International Atomic Energy Agency, Vienna, Austria.

ISO (International Organization for Standardization), 2001, Reference Neutron Radiations – Part 1: 

Characteristics and Methods of Production, ISO/CD 8529-1, Geneva, Switzerland.

Kathren, R. L., 1967, “Dosimetry of 14-MeV Neutrons with Eastman NTA Personnel Monitoring Film,” 

Health Physics, volume 13, number 9, pp. 1039–1040.

Kathren, R. L., 1987, "External Beta-Photon Dosimetry for Radiation Protection," The Dosimetry of 

Ionizing Radiation, Volume II, K. R. Kase and F. H. Attix, editors, pp. 321–370, Academic 

Press, New York, New York.

Kathren, R. L., 2004, Photon Doses and Energies at Selected Times following a Nuclear Detonation, 

The Kathren Group, Richland, Washington, July 28. [SRDB Ref ID: 13997]

Kathren, R. L., 2006, "Personal communication with R. Griffith" Oak Ridge Associated Universities 

Team, Oak Ridge, Tennessee.

Kathren, R. L., and H. V. Larson, 1969, “Radiological Calibration and Standardization for Health 

Physics: A Program, a Plea, and a Proposal,” Health Physics, volume 16, pp. 778–782. 

[SRDB Ref ID: 13356]

Kathren, R. L., C. T. Prevo, and S. Block, 1965, “Angular Dependence of Eastman Type A (NTA) 

Personnel Monitoring Film,” Health Physics, volume 11, number 10, pp. 1067–1069. [SRDB 

Ref ID: 6180]

Kathren, R. L., P. R. Zurakowski, and M. Covell, 1966, "Effect of Humidity and Dose on Personnel 

Monitoring Film," American Industrial Hygiene Association Journal, volume 27, pp. 388–395.

Kean, W. B., 1951, Exercise Desert Rock 1, Headquarters, III Corps, Fort MacArthur, California.

Kocher, D. C., and K. F. Eckerman, 1981, “Electron Dose-Rate Conversion Factors for External 

Exposure of the Skin,” Health Physics, volume 40, number 4, pp. 467–475.

Kocher, D. C., and K. F. Eckerman, 1987, “Electron Dose-Rate Conversion Factors for External 

Exposure of the Skin from Uniformly Deposited Activity on the Body Surface," Health Physics, 

volume 53, number 2, pp. 135–141.

Kulp, J. L., and J. L. Dick, 1960, “The Radiation Hazard from Contaminated Aircraft,” Health Physics, 

volume 4, pp. 133–156.

LANL (Los Alamos National Laboratory), 1994, "Final Hazard Analysis for Drillback Operations at the 

Nevada Test Site," memorandum, TPO-94-U-054, Los Alamos, New Mexico, February 14.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 71 of 121

Lehman, R. L., 1961, Energy Response and Physical Properties of NTA* Personnel Neutron 

Dosimeter Nuclear Track Film, UCRL-9513, University of California, Lawrence Radiation 

Laboratory, Berkeley, California, January 13. [SRDB Ref ID: 14860]

LRL (Lawrence Radiation Laboratory), 1964, "LRL Standard Operating Guide Postshot Drilling," 

memorandum, August 13.

LRL (Lawrence Radiation Laboratory), 1966, Postshot Drilling Handbook, UCID-15034, October.

Mayer, S., T. Otto, and N. Golnik, 2004, “Determination of the Photon-Contribution of a 238Pu-Be 

Source,” Nuclear Instruments and Methods in Physics Research, volume B213, pp. 214-217.

NCRP (National Council on Radiation Protection and Measurements), 1971, Protection Against 

Neutron Radiation, NCRP Report 38, Bethesda, Maryland, January 4.

NCRP (National Council on Radiation Protection and Measurements), 1978, A Handbook of 

Radioactivity Measurements Procedures, 2nd Edition, Report 58, Bethesda, Maryland.

NCRP (National Council on Radiation Protection and Measurements), 1990, Limit for Exposure to 

"Hot Particles" on the Skin, Report 106, Washington, D.C.

Nelms, A. T., and J. W. Cooper, 1959, “U-235 Fission Product Decay Spectra at Various Times After 

Fission,” Health Physics, volume 1, number 4, pp. 427–441.

NIOSH (National Institute for Occupational Safety and Health), 2006, External Dose Reconstruction 

Implementation Guideline, OCAS-IG-001, Rev. 2, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, August 25.

NRC (National Research Council), 1989, Film Badge Dosimetry in Atmospheric Nuclear Tests, 

National Academy of Sciences, National Academy Press, Washington, D.C. [SRDB Ref ID: 

1905]

NRDL (Naval Radiological Defense Laboratory), 1968, Hazards to Personnel Re-entering the Nevada 

Test Site Following Nuclear Reactor Tests, NRDL-TR-68-149, San Francisco, California.

ORAUT (Oak Ridge Associated Universities Team), 2004, Technical Basis Document for the Nevada 

Test Site – Site Description, ORAUT-TKBS-0008-02, Rev. 00, Oak Ridge, Tennessee, 

February 2.

ORAUT (Oak Ridge Associated Universities Team), 2005, Interpretation of Dosimetry Data for 

Assignment of Shallow Dose, ORAUT-OTIB-0017, Rev. 01, Oak Ridge, Tennessee, October 

11. 

ORAUT (Oak Ridge Associated Universities Team), 2006a, Technical Basis Document for the 

Nevada Test Site – Occupational Environmental Dose, ORAUT-TKBS-0008-04, Rev. 00 PC-1, 

Oak Ridge, Tennessee, December 8.

ORAUT (Oak Ridge Associated Universities Team), 2006b, Technical Basis Document for the 

Nevada Test Site – Occupational Internal Dose, ORAUT-TKBS-0008-5, Rev. 00 PC-1, Oak 

Ridge, Tennessee, July 20.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 72 of 121

ORAUT (Oak Ridge Associated Universities Team), 2006c, Pantex Plant National Security Complex – 

Occupational External Dose, ORAUT-TKBS-0013-6, Revision 00, Oak Ridge, Tennessee, July 

27.

ORAUT (Oak Ridge Associated Universities Team), 2006d, External Dose Reconstruction, ORAUTPROC-0006, Rev. 01, Oak Ridge, Tennessee, June 5.

ORAUT (Oak Ridge Associated Universities Team), 2007, Rocky Flats Plant – Occupational External 

Dosimetry, ORAUT-TKBS-0011-6, Rev. 00, Oak Ridge, Tennessee, February 8.

Pan Am (Pan American World Airways), 1967, Final Proposal for TLD System at NRDS for the Space 

Nuclear Propulsion Office, Nevada Extension, PAA 33-14, Pan American World Airways, 

September 15. [SRDB Ref ID: 13998]

REECo (Reynolds Electrical and Engineering Company), 1961, Standard Operating Procedures for 

the Radiological Safety Division, RRS-61-3, Nevada Test Site, Mercury, Nevada, January. 

[SRDB Ref ID: 1922]

REECo (Reynolds Electrical and Engineering Company), 1962, Basic Radiological Safety Training 

Manual, REECo-559:62-0140 (revision), Las Vegas, Nevada, January. [SRDB Ref ID: 1568]

REECo (Reynolds Electrical and Engineering Company), 1968, "Section 29: Personnel Dosimetry and 

Records," RadSafe Department Standard Operating Procedures, Las Vegas, Nevada. [SRDB 

Ref ID: 2007]

REECo (Reynolds Electrical and Engineering Company), 1985, “Determination of Radiokrypton and 

Radioxenon in Air,” Section 7, Environmental Sciences Department, Radioanalytical 

Laboratory Procedures, Mercury, Nevada. 

REECo (Reynolds Electrical and Engineering Company), 1990, Procurement and Acceptance 

Criteria, Standard Operating Procedure AABAC.D.02.40, Rev. 0, March 30. [SRDB Ref ID: 

2054, p. 131]

REECo (Reynolds Electrical and Engineering Company), 1995a, Nevada Test Site Radiation 

Protection Program, DOE/NV/11432-203, Rev. 3, Las Vegas, Nevada, December. [SRDB Ref 

ID: 14002]

REECo (Reynolds Electrical and Engineering Company), 1995b, Monthly Calibration of Gamma 

Radiation Fields at Inner and Outer Dosimeter Support Rings, C34-6209-086.A, Las Vegas, 

Nevada, November 30. [SRDB Ref ID: 14006]

REECo (Reynolds Electrical and Engineering Company), 1995c, Thermoluminescent Dosimetry 

Quality Assessment, C34-6209-039.A, Las Vegas, Nevada, November 30. [SRDB Ref ID: 

14007]

Roberts, R.B., 1937, “Investigation of the Deuteron-Deuteron Reaction”, Physical Review, volume 51, 

number 10, pp. 810–817.

Sanders, F. W., J. A. Auxier, and J. S. Cheka, 1959, “A Simple Method of Minimizing the Energy 

Dependence of Pocket Ionization Chambers,” Health Physics, volume 2, pp. 308–309.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 73 of 121

Schwartz, R. B., and C. M. Eisenhauer, 1980, The Design and Construction of a D2O-Moderator 

252Cf Source for Calibrating Neutron Personnel Dosimeters Used at Nuclear Power Reactors, 

NUREG-CR-1204, U.S. Nuclear Regulatory Commission, Washington, D.C.

Shipler, D. B., 2004, “Verification of Telephone Conversation with Dillard Shipler and Jerome Martin 

8/17/04,” letter to Robert Carr, Pacific Northwest National Laboratory, Richland, Washington, 

August 18.

Shipman, T. L., 1953, Operation Buster-Jangle Radiological Safety, WT-425(EX), University of 

California, Los Alamos Scientific Laboratory, Los Alamos, New Mexico, July (extracted 

version, October 1, 1979). [SRDB Ref ID: 2458]

Shipman, T. L., et al., 1951, "Log of JM-4 operations, rad-safe group, Operation RANGER, Site 

Monitoring, 23 January to 11 February, 1951," Nellis Air Force Base, Las Vegas, Nevada.

Shleien, B., L. A. Slaback, Jr., and B. K. Birky, editors, 1998, Handbook of Health Physics and 

Radiological Health, Third Edition, Williams & Wilkins, Baltimore, Maryland, January.

Sondhaus, C. A., and V. P. Bond, 1955, Physical Factors and Dosimetry in the Marshall Island 

Radiation Exposures, WT-939, U.S. Naval Radiological Defense Laboratory, San Francisco, 

California, December. [SRDB Ref ID: 14876]

Storm, E., 1951, memorandum to W. G. Kratz (Los Alamos Scientific Laboratory), Reynolds Electrical 

and Engineering Company, Las Vegas, Nevada, July 12.

Thierry-Chef, I., F. Pernicka, M. Marshall, E. Cardis, and P. Andreo, 2002, "Study of a Selection of 10 

Historical Types of Dosimeter: Variation of the Response to Hp(10) with Photon Energy and 

Geometry of Exposure," Radiation Protection Dosimetry, volume 102, number 2, pp. 101–113. 

[SRDB Ref ID: 11296]

Unger, L. M., and D. K. Trubey, 1981, Specific Gamma-Ray Dose Constants for Nuclides Important to 

Dosimetry and Radiological Assessment, ORNL/RSIC-45, Oak Ridge National Laboratory, 

Oak Ridge, Tennessee.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 74 of 121

GLOSSARY

absorbed dose, D

Amount of energy imparted by radiation to unit mass of absorbing material (100 ergs per 

gram), including tissue. The unit used before the use of the International System of metric 

units is the rad; the International System unit is the gray.

accreditation

Recognition that a dosimeter system has passed the performance criteria of the DOE 

Laboratory Accreditation Program standard in specified irradiation categories.

accuracy

If a series of measurements has small systematic errors, they are said to have high accuracy. 

The accuracy is represented by the bias.

albedo dosimeter

A TLD device that measures the thermal, intermediate, and fast neutrons that are scattered 

and moderated by the body from an incident fast neutron flux.

algorithm

A computational procedure.

Atomic Energy Commission

Original agency established for nuclear weapons and power production; a successor to the 

Manhattan Engineer District , and a predecessor to the U.S. Department of Energy.

backscatter

Deflection of radiation by scattering processes through angles greater than 90 degrees with 

respect to the original direction of motion.

beta particle

A charged particle of very small mass emitted spontaneously from the nuclei of certain 

radioactive elements. Most (if not all) direct fission products emit (negative) beta particles. 

The beta particle is physically identical with an electron moving at high velocity.

buildup

Increase in flux or dose due to scattering in the medium.

calibration blank

A dosimeter that has not been exposed to a radiation source. The results from this dosimeter 

establish the dosimetry system baseline or zero dose value.

collective dose equivalent

The sum of the dose equivalents of all individuals in an exposed population. Collective dose is 

expressed in units of person-rem (person-sievert).

control dosimeter

A dosimeter used to establish the dosimetry system response to radiation dose. The 

dosimeter is exposed to a known amount of radiation.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 75 of 121

curie

A special unit of activity. One curie exactly equals 3.7 × 1010 nuclear transitions per second.

deep absorbed dose (Dd) 

The absorbed dose at the depth of 1.0 centimeter in a material of specified geometry and 

composition.

deep dose equivalent (Hd) 

The dose equivalent at the respective depth of 1.0 centimeter in tissue.

densitometer

Instrument that has a photocell to determine the degree of darkening of developed 

photographic film.

density reading

See optical density.

DOE Laboratory Accreditation Program (DOELAP)

Accredits DOE site dosimetry programs based on performance testing and onsite reviews 

performed on a 2-year cycle.

dose equivalent (H) 

The product of the absorbed dose D, the quality factor Q, and any other modifying factors. 

The special unit is the rem. When D is expressed in gray, H is in sieverts. (1 sievert = 

100 rem.)

dosimeter

A device used to measure the quantity of radiation received. A holder with radiation-absorbing 

elements (filters) and an insert with radiation-sensitive elements packaged to provide a record 

of absorbed dose or dose equivalent received by an individual. (See albedo dosimeter, film 

dosimeter, neutron film dosimeter, thermoluminescent dosimeter.)

dosimetry system

A system used to assess dose equivalent from external radiation to the whole body, skin, and 

extremities. This includes the fabrication, assignment, and processing of dosimeters as well 

as interpretation and documentation of the results.

DuPont 552

A film packet containing two pieces of film: a 502 sensitive film and a 510 insensitive film.

DuPont 558

A film packet containing a 508 film with one side having a sensitive emulsion and the other 

side insensitive emulsion.

error

A term used to express the difference between the estimated and actual value. 

exchange period (frequency)

Period (weekly, biweekly, monthly, quarterly, etc.) for routine exchange of dosimeters.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 76 of 121

exposure

As used in the technical sense, a measure expressed in roentgens of the ionization produced 

by gamma (or X-) rays in air. 

fast neutron

Neutron of energy between 10 kilovolts-electron and 10 megavolts-electron.

favorable to claimants

Process of estimation based on technical considerations of parameters significant to dose 

such that the estimated dose is not underestimated.

film

Generally means a film packet that contains one or more pieces of film in a light-tight 

wrapping. When developed the film has an image caused by radiation that can be measured 

using an optical densitometer. (See DuPont 552, DuPont 558, nuclear track emulsion)

film density

See optical density.

film dosimeter

A small packet of film in a holder that attaches to a worker.

filter

Material used to adjust radiation response of a dosimeter to provide an improved tissue 

equivalent or dose response.

first collision dose

A dose measurement that can be determined for photons or neutrons. For neutron radiation, 

the simplest calculation is one relating dose to flux through a thin layer of tissue. The resultant 

graph, sometimes referred to as the first-collision curve, is derived from the assumption that 

the probability of two or more interactions per neutron is negligible. Because of the short 

range of the charged secondary radiation from fast neutrons, the first collision dose in 

irradiated material is practically the same as the absorbed dose.

gamma rays

Electromagnetic radiation (photons) originating in atomic nuclei and accompanying many 

nuclear reactions (e.g., fission, radioactive decay, and neutron capture). Gamma rays are 

physically identical to X-rays of high energy; the only essential difference is that X-rays do not 

originate in the nucleus. 

gamma ray interactions

Interaction of gamma rays with matter occurs through three primary processes as follows:

• Photoelectric absorption

The process whereby a gamma ray (or X-ray) photon, with energy somewhat greater than 

that of the binding energy of an electron in an atom, transfers all its energy to the electron, 

which is consequently removed from the atom.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 77 of 121

• Compton scattering 

An attenuation process observed for X-ray or gamma radiation in which an incident photon 

interacts with an orbital electron of an atom to produce a recoil electron and a scattered 

photon of energy less that the incident photon.

• Pair production

An absorption process for X-ray and gamma radiation in which the incident photon is 

annihilated in the vicinity of the nucleus of the absorbing atom, with subsequent production 

of an electron and positron pair. This reaction occurs only for incident photon energies that 

exceed 1.02 megavolts-electron.

gray

The International System unit of absorbed dose (1 gray = 100 rad).

intermediate-energy neutron

Neutron of energy between 0.5 electron volt (assumed to be 0.4 electron volt because of 

cadmium cutoff in neutron response) and 10 kilovolts-electron.

ionizing radiation

Electromagnetic radiation (consisting of photons) or particulate radiation (consisting of 

electrons, neutrons, protons, etc.) capable of producing charged particles through interactions 

with matter.

isotopes

Forms of the same element having identical chemical properties but different atomic masses. 

Isotopes of a given element all have the same number or protons in the nucleus but different 

numbers of neutrons. Some isotopes of an element can be radioactive.

kilovolt-electron (keV)

An amount of energy equal to 1,000 electron volts.

nuclear track emulsion, type A (NTA)

A film that is sensitive to fast neutrons. The developed image has tracks caused by neutrons 

that can be seen by using oil immersion and 1,000-power microscope.

luminescence

The emission of light from a material as a result of some excitation.

Manhattan Engineer District

Agency designated to develop nuclear weapons; a predecessor to the U.S. Department of 

Energy.

minimum detection level (MDL)

Often confused because statistical parameters necessary to its calculation are not explicitly 

defined. Nonetheless, the MDL is often assumed to be the level at which a dose is detected at 

the 2-sigma level (i.e., 95% of the time). The MDL should not be confused with the minimum 

recordable dose. 

megavolt-electron (MeV)

An amount of energy equal to 1 million electron volts.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 78 of 121

multiple-collision neutron dose

Dose to flux through tissue based on the assumption that two or more interactions per neutron 

occur resulting in greater energy deposition. 

nuclear emulsion

Generally refers to Nuclear Track Emulsion, type A film.

neutron

A basic particle that is electrically neutral weighing nearly the same as the hydrogen atom.

neutron film dosimeter

A film dosimeter that contains a nuclear track emulsion, type A film packet.

nonpenetrating dose (NP or NPen)

Designation on film dosimeter reports that implies a radiation dose, typically to the skin of 

whole body, from beta and lower energy photon radiation.

open window (OW)

Designation on Hanford film dosimeter reports that implies the use of little shielding (only that 

of the security credential). It commonly is used to label the film response corresponding to the 

OW area. 

optical density

The quantitative measurement of the density of photographic blackening defined as 

D = Log10(Io/I).

pencil dosimeters

A type of ionization chamber used by personnel to measure radiation dose. These results 

might be labeled as Pen dose. Other names: pencil, pocket dosimeter, pocket pencil, pocket 

ionization chamber, PIC.

penetrating dose (P or Pen)

Designation on film dosimeter reports that implies a radiation dose, typically to the whole body, 

from higher energy photon radiation.

personal dose equivalent, HP(d)

Radiation quantity recommended for use as the operational quantity to be recorded for 

radiological protection purposes by the International Commission on Radiological Units and 

Measurements. Represented by HP(d), where d identifies the depth (in millimeters) and 

represents the point of reference for dose in tissue. For weakly penetrating radiation of 

significance to skin dose, d = 0.07 millimeter and is noted as HP(0.07). For penetrating 

radiation of significance to whole-body dose, d = 10 millimeters and is noted as HP(10). 

photon

A unit or particle of electromagnetic radiation consisting of X- and/or gamma rays. 

quality factor (Q)

A modifying factor used to derive dose equivalent from absorbed dose.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 79 of 121

rad

A unit of absorbed dose equal to the absorption of 100 ergs per gram of absorbing material, 

such as body tissue.

radiation

One or more of beta, neutron, and photon radiation. 

radiation monitoring

Routine measurements and the estimation of the dose equivalent for the purpose of 

determining and controlling the dose received by workers.

radioactivity

The spontaneous emission of radiation, generally alpha or beta particles, gamma rays, and 

neutrons from unstable nuclei.

random errors

When a given measurement is repeated the resultant values, in general, do not agree exactly. 

The causes of the disagreement between the individual values must also be causes of their 

differing from the actual value. Errors resulting from these causes are called random errors.

relative biological effectiveness (RBE)

A ratio of the absorbed dose of a reference radiation to the absorbed dose of a test radiation 

producing the same biological effects, other conditions being equal.

rem

The rem is a unit of dose equivalent, which is equal to the product of the number of rads 

absorbed and the quality factor. The word derives from roentgen equivalent in man.

rep

Historically the rep has been used extensively for the specification of permissible doses of 

ionizing radiations other than X-rays or gamma rays. Several definitions have appeared in the 

literature but in the sense most widely adopted, it is a unit of absorbed dose with a magnitude 

of 93 ergs per gram. The word derives from roentgen-equivalent-physical.

roentgen

A unit of exposure to gamma (or X-ray) radiation. It is defined precisely as the quantity of 

gamma (or X-) rays that will produce a total charge of 2.58 × 10-4 coulomb in 1 kilogram of dry 

air. An exposure of 1 roentgen is approximately equivalent to an absorbed dose of 1 rad in 

soft tissue.

scattering

The diversion of radiation from its original path as a result of interactions with atoms between 

the source of the radiations and a point at some distance away. Scattered radiations are 

typically changed in direction and of lower energy than the original radiation.

shallow absorbed dose (Ds)

The absorbed dose at a depth of 0.07 millimeter in a material of specified geometry and 

composition.

shallow dose equivalent (Hs)

Dose equivalent at a depth of 0.07 millimeter in tissue.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 80 of 121

shielding

Any material or obstruction that absorbs (or attenuates) radiation and thus tends to protect 

personnel or materials from radiation.

sievert

The International System unit for dose equivalent. (1 sievert = 100 rem.)

skin dose

Absorbed dose at a tissue depth of 7 milligrams per square centimeter.

thermal neutron

Strictly, neutrons in thermal equilibrium with surroundings. Generally, refers to neutrons of 

energy less than the cadmium cutoff of about 0.4 electron volt.

tissue-equivalent

Used to imply that radiation response characteristics of the material being irradiated are 

equivalent to tissue. Achieving a tissue-equivalent response is an important consideration in 

the design and fabrication of radiation measuring instruments and dosimeters.

thermoluminescent

Property of a material that causes it to emit light as a result of being excited by heat.

thermoluminescent dosimeter (TLD)

A holder containing solid chips of material that when heated will release the stored energy as 

light. The measurement of this light provides a measurement of absorbed dose. The solid 

chips are sometimes called crystals.

thermoluminescent dosimeter chip

A small block or crystal made of LiF used in the TLD.

TLD-600 - A TLD chip made from 6

Li (>95%) used to detect neutrons.

TLD-700 - A TLD chip made from 7

Li (>99.9%) used to detect photon and beta radiation.

whole-body dose

Absorbed dose at a tissue depth of 1.0 centimeter (1,000 milligrams per square centimeter); 

however, also used to refer to the dose recorded.

X-ray

Ionizing electromagnetic radiation of extranuclear origin.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 81 of 121

TABLE OF CONTENTS

SECTION TITLE

A.1 Beta/Gamma Dosimetry.........................................................................................................

PAGE

81

A.1.1 Photographic Film Dosimeters....................................................................................81

A.1.2 Thermoluminescent Dosimeters .................................................................................84

A.1.2.1 NRDS Thermoluminescent Dosimeters........................................................ 84

A.1.2.2 REECo Thermoluminescent Dosimeters ...................................................... 85

A.1.2.3 Energy Response......................................................................................... 87

A.1.2.4 Dose Measurement...................................................................................... 88

A.1.2.5 Calibration.................................................................................................... 88

A.2 Neutron Dosimetry ................................................................................................................. 89

A.2.1 NTA Film ....................................................................................................................89

A.2.2 TLD Albedo Detectors ................................................................................................91

A.2.3 Track Etch Detectors ..................................................................................................92

A.2.3.1 Dose Algorithms........................................................................................... 93

A.3 Quality Assurance..................................................................................................................94

A.4 Beta Particle Field Characteristics.......................................................................................... 95

A.5 Neutron Field Characteristics ................................................................................................. 97

A.1 BETA/GAMMA DOSIMETRY

A.1.1

Photographic emulsions of various types and in various holders (i.e., film badges) were used at NTS 

for personnel monitoring from the start of operations in January 1951 until 1987 (DeMarre 2002; Allen 

and Schoengold 1995). The basic principles, theory, and practice of photographic film dosimetry for 

beta and photon radiation are well known and have been described in a number of standard texts and 

references (Becker 1966; Dudley 1966; Ehrlich 1954, 1962; Kathren 1987; NRC 1989). The following 

paragraphs discuss factors related to film dosimetry that provide a general background to the 

interpretation and reconstruction of dosimetry results at NTS.

Photographic Film Dosimeters

The term film badge, or film badge dosimeter, as used in this document, refers to the entire dosimeter 

issued to personnel, which typically consisted of a dental-size film packet housed in a holder of 

varying sophistication designed to improve the response characteristics and measurement 

capabilities. The film consisted of a plastic base covered on one or both sides with a layer of a 

suitable photographic emulsion. One or two pieces of film were wrapped in a light-tight paper 

package to comprise the packet, which in turn was placed in a holder containing metallic filter(s) to 

compensate for the photon energy dependence of the film. The sensitivity of the film to ionizing 

radiation was largely a function of the size of the AgBr grains in the emulsion. Because the typical 

ATTACHMENT A

DOSIMETRY TECHNOLOGY

Page 1 of 20
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 82 of 121

photographic emulsion for dosimetry has an effective range of about 3 orders of magnitude, the 

packet typically contained two pieces of film with different sensitivities. The effective ranges of these 

films overlapped, which permitted an effective response range of about 5 orders of magnitude for 

penetrating photon radiations. 

In principle, film dosimetry is simple; it consists of the determination of optical density or degree of 

blackening produced by exposure to radiation. The degree of blackening produced by the radiation 

incident on the film is determined in terms of the net optical density, the logarithm of which is typically 

plotted against the logarithm of exposure or dose to produce a calibration or response curve, or an 

algorithm with which the exposure registered by the film can be determined. The response of a film 

emulsion to photon radiation is not linear with dose or exposure. Near the lower (less than 50 mR) 

and upper (about 1.5 to 5 R) limits of the range of the film, small increments of density represent 

relatively large changes in dose. Therefore, the uncertainty at the low and high ends of the dose 

response curve (i.e., net optical density vs. exposure), where the slope is quite shallow, is relatively 

large. However, the total uncertainty for high-energy exposures (more than about 150 keV to 3 MeV) 

for these dose ranges (1.5 to 5 R) should not exceed a factor of 2. For high-energy photon exposures

in the region of 50 to 1,500 mR (the steep portion of the dose response curve), the uncertainty is 

much less and should be within 20% to 30%. Dose reconstructors should consider these uncertainty 

values to be broad estimates, not exact values.

Although a number of different film types were used for dosimetry at NTS, they generally had similar 

characteristics and responses to beta and photon radiations. Uncertainties in dosimetry with these 

films are largely, if not exclusively, the result of external factors rather than differences in the films.

The response of film to photon radiation is affected by a number of interrelated variables. From the 

standpoint of dose reconstruction at NTS, the most important of these are energy dependence, 

angular dependence, and effects of temperature and humidity. Because of the high relative atomic 

number (Z) of the AgBr in the film emulsion in relation to soft tissue, the response or energy absorbed 

per unit exposure of the film is a strong function of the energy of the exposing photons, rising steeply 

at energies below about 200 keV to a peak at about 30 keV, and then falling off steeply. At about 

30 keV, the energy of maximum response, the degree of blackening per unit exposure is about 30-fold 

greater than that for photons with energies of few hundred thousand to a few million electron volts. 

Because the relationship between exposure and soft tissue dose is approximately constant over a 

wide energy range, it was essential to compensate or correct for the energy dependence of the 

photographic response. This was accomplished by placing metallic filters over portions of the film to 

alter or flatten the energy dependence characteristics such that the film responded more like soft 

tissue over a wide energy range.

Although energy dependence can be a large source of error in film badge dosimetry, reasonably good 

results can be obtained by a four-element badge: one with an OW or no filter over a portion of the film 

packet, and high-Z, medium-Z, and low-Z filters over other portions of the packet. By observing the 

relationships of the degree of darkening under each filter area, and comparing these with exposures 

to known energies and doses of photon radiations, a reasonably accurate assessment of photon dose 

can be made over a wide range of energies. This is particularly true at NTS, where exposures were 

largely to photons with energies greater than a few hundred thousand electron volts at angles close to 

normal to the plane of the film. In general, errors in dose interpretation that result from energy 

dependence will result in overestimates of dose and are, therefore, conservative and favorable to 

ATTACHMENT A

DOSIMETRY TECHNOLOGY

Page 2 of 20
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 83 of 121

claimants. However, given the single-element badges in use until 1966, some dose from lower 

energy photons, specifically those with energies below about 80 keV, could have been missed 

because of attenuation in the filter; therefore, the dose interpretation could be low. Given the photon 

energy spectra at NTS, any such loss of dose is likely to have been relatively low.

Photon calibration of film dosimeters was typically performed in terms of exposure in free air, with the 

plane of the film oriented normally to the direction of the incident photon beam. Initial calibrations 

were made with a high-energy photon source (e.g., 60Co or 137Cs) with an energy response by film 

similar to the energy spectra to which personnel were exposed in the field. After introduction of the 

advanced multielement film badge in January 1966, calibration was conducted with X-ray sources as 

well as high-energy photons. This enabled interpretation of exposure to the film with a higher degree 

of accuracy over a broader spectrum of photon energies.

The response of film, whether in a bare film packet or in a holder, is a function of the angle of 

incidence of the exposing radiation on the plane of the film. Because the film is in a small flat sheet, 

an edge-on exposure (i.e., an exposure with the incident radiation parallel to the plane of the film) 

produces a different effect or optical density pattern than an exposure normal (at right angles) to the 

plane of the film. The effect is strongly dependent on both the energy of the exposing photons and 

the type of film holder or film badge, but is generally minimal at photon energies above a few hundred 

thousand electron volts and for angles of incidence ranging from about 30 to 150 degrees in relation 

to the plane of the film. However, for angles of incidence approaching parallelism (i.e., 0 degrees) 

with the plane of the film, the effect can be pronounced, and can lead to significant underestimates in 

dose. The problem, however, should be more or less minimal for exposures at NTS because these 

were (1) typically to high-energy photons and at angles close to normal with the plane of the film and 

(2) probably largely accounted for in the interpretation of the badge.

Environmental conditions can significantly affect film badge results, and can result in large 

uncertainties and errors in dose estimation. Temperatures greater than 130o

F (50oC) can induce 

fogging, but several days at this temperature are necessary before fogging or increased density 

occurs. At higher temperatures, which could have occurred in closed vehicles or buildings at NTS, 

fogging can result after shorter exposure times (Kathren, Zurakowski, and Covell 1966). If not 

corrected, high-temperature fogging always produces an overestimate and, therefore, a dose 

interpretation favorable to claimants. 

High relative humidity results in a fading of the latent image before development and in a decrease in 

the measured density of the film for a given dose. Latent image fading results in a low estimate of 

dose. However, studies have shown that latent image fading is not a problem until films have been 

worn for intervals exceeding 4 to 6 wk, and can be largely overcome by encasing the film packet in a 

polyethylene pouch, as was done for hot and humid conditions encountered during nuclear tests in 

the Pacific (Kathren, Zurakowski, and Covell 1966). Given the low-humidity conditions at NTS, latent 

image fading is unlikely to be of significance to film dosimetry.

Films were exchanged at varying intervals and, once collected, were developed under controlled 

conditions. Optical densities were determined with a densitometer, and doses were determined from 

calibration curves, which were log-log plots of net optical density versus dose obtained from a series 

of films exposed to known levels of photons. The use of control films at sites such as the badge 

house, where there was no expectation of a radiation field other than from background, permitted the 

ATTACHMENT A

DOSIMETRY TECHNOLOGY

Page 3 of 20
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 84 of 121

inherent density of the film and any density attributable to background radiation (the so-called 

background fog) to be subtracted from the reading, thereby providing the net optical density 

attributable to occupational exposure. In practice, background films were used to zero the 

densitometer, which then gave a net optical density reading. 

A.1.2

A.1.2.1 NRDS Thermoluminescent Dosimeters

Thermoluminescent Dosimeters

The first routine use of TLDs at NTS began in 1970 (Boone, Bennett, and Adams 1970). Starting in 

February 1966, Pan Am used TLDs at the NRDS as part of the site effluent monitoring program. 

These dosimeters contained a calcium-fluoride phosphor bound to a helically wound wire in an 

evacuated glass tube and were ideal for the intended purpose but unsuitable for personnel dosimetry

(Boone, Bennett, and Adams 1970). 

Other installations had successfully demonstrated the applicability of LiF TLDs for personnel 

dosimetry (Boone, Bennett, and Adams 1970). A brief investigation was performed and proposals 

submitted for institution of a LiF TLD program at NRDS in 1967 to complement the REECo-supplied 

personnel film dosimetry service.

On July 1, 1968, Pan Am initiated a TLD program at the NRDS with the cooperation of the Health 

Services Laboratory of the AEC Idaho Operations Office (IDO). This laboratory supplied LiF TLD 

chips and provided readout services and considerable consultation based on its extensive LiF TLD 

experience at the National Reactor Testing Station. From July 1, 1968, until July 1970 NRDS 

personnel who were not likely to receive a significant exposure received TLDs. All reported 

exposures were in the background range (0 to 30 mrem). Routine use of TLDs began in July 1970 

and continued until the NRDS ceased operation at the end of 1972 (Boone, Bennett, and Adams

1970; DOE 2002). 

Beginning in July 1970 (and repeated every quarter thereafter), a minimum of one set of dosimeters 

was exposed on the Pan Am Calibration Range to a total dose(s) of from 100 mR to 5 R. Each set of 

dosimeters consisted of 1 Pan Am R-chamber, 6 REECo film, and 12 IDO TLD chips (6 to be read out 

at Pan Am and 6 at IDO). The Pan Am Calibration Range was crosschecked against Pan Am 

R-chambers, which were calibrated by NBS; (one set was sent to NBS annually). If the difference 

between the range and the R-chamber exceeded ±6%, the data points were retaken (Boone, Bennett, 

and Adams 1970). 

A calibration factor was calculated for each type of dosimeter as follows:

R

DoDx CF − = (A-1)

where:

CF = calibration factor

Dx = mean indicated dose from the set of six exposed dosimeters

Do = mean indicated dose from the set of six unexposed dosimeters

ATTACHMENT A

DOSIMETRY TECHNOLOGY

Page 4 of 20
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 85 of 121

R = R-chamber indicated dose.

As long as the average calibration factor over the dose range of interest fell in the range of 0.9 to 1.1 

(±10%), the results were documented but no further action was taken because this level of accuracy 

was considered adequate. If the average calibration factor fell outside this range, the parties 

concerned were notified and another set of data was taken. In general, the relative accuracy of both 

measurement systems, as indicated by the body of data considered, is adequate for personnel 

dosimetry purposes.

In the case of Pan Am-read TLDs, a calibration factor was always necessary due to the relative 

response of the reader. The appropriate calibration factor was employed as follows:

( ) −×= DoDiCFDc (A-2)

where:

Dc = calibration factor

Do = actual dose

Di = indicated dose

Boone, Bennett, and Adams (1970) contains additional information on calibrations and calibration 

comparisons between Pan Am, IDO, and REECo.

Figure 6-2 shows the security badge and enclosed TLD insert. It was a plastic holder in which two LiF 

dosimeters were placed. One dosimeter was shielded by the tantalum-cadmium filter area of the 

present badge. The other dosimeter was unshielded. The two dosimeters were held in place by 

retainers built into the insert (for pliable LiF-Teflon discs) or hinged retainers (for extruded LiF chips). 

The insert was punched with a binary code that identified security badges for issuance and collection. 

Finally, the bottom left-hand portion of the dosimeter insert was coded to enable security guards to 

determine if dosimeters were being worn for the proper period.

A.1.2.2 REECo Thermoluminescent Dosimeters

With the advent of DOE requirements to restrict personnel exposures to ALARA and with emphasis 

on accurate dosimetry at low doses, REECo Environmental Sciences Department personnel began 

evaluating TLD systems and neutron dosimeters in the early 1980s to replace the film badge and 

neutron TLD (Allen and Schoengold 1995). 

After evaluating several dosimetry systems, the Environmental Sciences Department determined that 

the Panasonic 802 TLD and the neutron TED were the best combination for NTS exposure conditions. 

These were put into use January 1, 1987. The security credential holder was redesigned to 

accommodate both dosimeters (Figures 6-3 and 6-4).

DOE Order 5480.11, "Radiation Protection for Occupational Workers” (DOE 1988), established the 

radiation protection standards and program for DOE, its contractor personnel, and other occupational 

personnel to protect workers from ionizing radiation. It defined policies and procedures required to 

operate DOE facilities and conduct activities to keep personnel exposures well below the limits set by 

ATTACHMENT A

DOSIMETRY TECHNOLOGY

Page 5 of 20
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 86 of 121

Order 5480.11 to meet ALARA goals. The site-specific Nevada Operations Office/Yucca Mountain 

Project (YMP) Radiological Control Manual (known as the RadCon Manual; DOE 1994) describes 

radiation protection standards and program requirements as they relate to the NTS and YMP 

organizations. The Nevada Test Site Radiation Protection Program, published in May 1995 (REECo 

1995a), demonstrated compliance with DOE rules for protecting individuals from ionizing radiation.

The four-element Panasonic UD-802 TLD was the primary dosimeter for routine use issued to all 

monitored personnel until 2001, when it was replaced by the Panasonic 809 dosimeter, which also 

contained four elements (Table 6-1). The Panasonic UD-807 TLD is used to monitor personnel 

working in situations where the likelihood of exposure to an extremity is significantly greater than 

exposure to the whole body.

In most instances, dosimeters (i.e., TLDs) were processed quarterly. Some personnel working in 

locations where high exposures were more likely to occur exchanged their dosimeters monthly. The 

NTS monitoring program was designed to ensure that personnel exposures were kept below the 

annual limit of a total effective dose equivalent of 5 rem.

The Panasonic UD-802 TLD was designed to identify the type and energy of detected radiation, so 

the prescribed tissue depth dose equivalents could be determined accurately. In specific, the fourelement Panasonic 802 was intended to measure the following:

• External photon radiation from 0.010 to 1,000 rem

• Gamma energy range of 0.010 to 10 MeV

• Beta radiation from 0.030 to 1,000 rad

• Beta energy range from approximately 0.30 to 10 MeV

Table A-1 lists TLD element composition, filtration, and radiation type.

Table A-1. Panasonic UD-802 dosimeter configuration (Allen and Schoengold 1995). 

Element Phosphor Filtration

Rad. type & dose 

equivalent

E1 Li2B4O7:Cu Teflon/polyester, 18 mg/cm2 Gamma, beta, shallow dose

E2 Li2B4O7:Cu Teflon/polyester and plastics, 70 mg/cm2 Gamma, beta, eye dose

E3 CaSO4:Tm Teflon/polyester, ABS plastics, and Black ceramic, 

645 mg/cm2

Gamma

E4 CaSO4:Tm Teflon/polyester, ABS and lead, 1,042 mg/cm2 Gamma, deep dose

The dosimeters were calibrated against known exposures to provide an accurate transition from 

measured exposure to dose equivalent. Dosimeter calibration factors were normalized to the 

corrected readings from run calibration factor (RCF) dosimeters processed with the field dosimeters. 

The computer program separated the calibration dosimeters, used the known RCF exposure value, 

and calculated RCFs that were applied to the remainder of the dosimeters in the run. The RCFs kept 

a reader in calibration over long periods and maintained consistency among different readers (Allen 

and Schoengold 1995). 

ATTACHMENT A

DOSIMETRY TECHNOLOGY

Page 6 of 20
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 87 of 121

A.1.2.3 Energy Response

Panasonic UD-802 and UD-809 TLD elements have well-known energy responses (Allen and 

Schoengold 1995) [39]. Like many radiation detection devices, there is an energy (about 20 keV) 

below which radiation does not deliver sufficient energy to the TLD element for detection. As photon 

radiation energy increases, the Li2B4O7:Cu elements have a relatively flat response, while the 

CaSO4:Tm elements over-respond between about 50 and 200 keV. The higher CaSO4:Tm response 

is due primarily to the effective Z of 14.4 for CaSO4, compared to 7.3 for Li2B407:Cu. The lithium 

borate effective Z is much closer to the effective Z for tissue (7.4), and this relationship is important to 

photoelectric effect interactions. The response of both types of elements is relatively flat to about 5 

MeV, where the CaSO4 response increases slightly due to its higher effective Z and, in this case, pair 

production interactions. Figure A-1 shows typical photon energy response curves for Li2B4O7 and 

CaSO4 TLD elements. Figure A-2 shows the response for the UD-802 TLD, including filters. The first 

element in the UD-802 TLD has a total filtration density depth of about 18 mg/cm2

, making it sensitive 

to beta particle radiation with energies of 100 keV or more. The second element responds to beta 

radiation with energies above about 300 keV. Because beta particles have a low linear energy 

transfer and the Li2B4O7 is nearly equivalent to tissue, the beta energy response is relatively flat and 

similar to the photon response shown in Figure A-2 for photon radiation (starting at 0.1 MeV for 

element E1 and 0.3 MeV for element E2). 

Figure A-1. Typical photon energy response curves for 

Li2B4O7 and CaSO4 TLD elements (Allen and 

Schoengold 1995).

NTS algorithms identify radiation types and energies and then apply factors necessary to obtain a 

correct dose equivalent from a TLD reading. Thus, the energy response characteristics of the UD-802 

TLD, over the useable range of the TLD, are automatically factored into the dose equivalent 

determination. The validity of the algorithms for energy response correction is verified through the 

satisfactory completion of DOELAP accreditation for photon and beta particle radiation in several 

different radiation energy categories (Allen and Schoengold 1995).

Unlike the Panasonic UD-802, the UD-809 introduced in 2001 has four lithium borate (Li2B4O7(Cu)). 

One uses lithium, enriched to 99.99% in 7

Li and 11B, both of which have a negligible neutron 

response. The other three employ neutron-sensitive 6

Li2

10B4O7(Cu) chips, with 6

Li enriched to 95.33% 

and 10B enriched to 94.64%. These two isotopes have high n,alpha cross sections. The elements are 

shielded with tin and cadmium on front and back, in various combinations [40].

ATTACHMENT A

DOSIMETRY TECHNOLOGY

Page 7 of 20
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 88 of 121

Figure A-2. Panasonic UD-802 energy 

responses (Allen and Schoengold 

1995).

A.1.2.4 Dose Measurement

The background, element correction factor (ECF), and RCF corrected measurements from TLD 

elements provide an indication of the dose received by the person wearing the dosimeter. However, 

because TLDs are not truly tissue equivalent in relation to dose equivalent, algorithms are used to 

convert the dosimeter response to a dose equivalent value at the specific depth in tissue (Allen and 

Schoengold 1995). 

The NTS algorithms were developed by irradiating TLDs mounted on a phantom of tissue substitute to 

specific radiation quantities and types (Allen and Schoengold 1995). The dosimeter measurement 

was compared to the dose equivalent calculated by knowing the type and energy of the radiation. 

This process was repeated many times until enough data were collected to develop equations and 

relationships between a dosimetric reading and the dose equivalent at the specified depth in tissue. 

Published factors for converting radiation exposure to dose equivalent were used in the development 

of the algorithms. Table A-2 lists exposure to dose equivalent conversion factors specified in the 

DOELAP Standard for NIST reference photon fields.

A.1.2.5 Calibration

Several Quality Assurance (QA) operations are performed on new dosimeters before they are 

approved for field use. The numerical coding on each TLD label is verified to ensure accurate 

tracking. Before use, and annually thereafter, dosimeter phosphor elements are calibrated by 

determining an ECF for each element. In accordance with applicable procedures, a dosimeter with an 

ECF out of tolerance is removed from use.

TLDs are calibrated with the NTS 137Cs gamma source calibration range. The 137Cs source is 

traceable to NIST through a secondary standard calibrated ion chamber. An integrating electrometer, 

calibrated annually for NIST traceability, is used to verify the accuracy of each TLD and TLD reader 

ATTACHMENT A

DOSIMETRY TECHNOLOGY

Page 8 of 20
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 89 of 121

calibration exposure. All temperature and barometric pressure measuring instruments used for the 

calibration range are calibrated annually to NIST traceable standards.

Table A-2. Photon exposure to dose conversion 

factors for NIST reference radiationsa

. 

NIST reference 

radiation

Conversion factor (mrem/R)

Shallow (0.07 mm) Deep (10 mm)

Filtered X-rays

M30 1.08 0.45

S60 1.15 1.07

M150 1.41 1.47

H150 1.41 1.41

K-fluorescence X-ray (keV)

16 1.08 0.38

24 1.07 0.47

34 1.07 0.99

43 1.28 1.30

58 1.47 1.54

78 1.61 1.72

100 1.59 1.74

Cesium-137 photon (keV)

662 1.03 1.03

a. Source: Allen and Schoengold (1995). 

Monthly Calibration of Gamma Radiation Fields at Inner and Outer Dosimeter Support Rings (REECo 

1995b) outlines procedures for performing gamma source exposures. The results of the exposures 

are used to ensure proper performance of the gamma sources, provide required exposure rate 

information to upgrade source calibration, and document exposure trends for quality control (QC).

A.2 NEUTRON DOSIMETRY

A.2.1

Fast neutrons interact with the hydrogen in the NTA emulsion and film, producing recoil protons that, 

having kinetic energy and being charged, travel through the emulsion, creating a string of exposed 

individual grains of AgBr along their path (Lehman 1951). On development, these show up as tracks 

of grains of developed AgBr. Track length is a function of the energy of the recoil proton and the 

angle of travel with respect to the plane of the film. Fast neutron dose is determined by direct visual 

counting, usually via a microscope, of the number of proton recoil tracks in a predetermined (for 

statistical purposes) number of microscope fields, usually 100. A countable track must have a length 

of at least three grains (i.e., there must be three grains in a row for a track to be registered).

NTA Film

Tracks can be produced in NTA film emulsion either by protons produced by the 14N(n,p) reaction with 

low-energy neutrons (less than 10 eV) or by direct recoils from energetic neutron interactions with 

hydrogen in the film. The relatively low thermal neutron fluences, the low (n,p) reaction cross-section, 

and the large fluence of thermal neutrons per millirem compared to fast neutrons (2,200 compared to 

7.5) rendered thermal neutron dosimetry impractical.

ATTACHMENT A

DOSIMETRY TECHNOLOGY

Page 9 of 20
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 90 of 121

Based on theoretical considerations, a minimum neutron energy of about 450 keV is needed to 

produce a proton recoil track, although in practice the minimum detectable energy or threshold energy 

was about 800 keV (Figure A-3). Thus, NTA film is essentially insensitive to neutrons below 500 to 

800 keV, depending on the quality of processing. Therefore, it is suitable only for occupational 

environments in which the majority of the dose comes from neutrons with energies above 1 MeV, or in 

which the fraction of neutrons below 1 MeV is reasonably well known. It is important to calibrate NTA 

film dosimeters with a source having a neutron energy spectrum similar to that in which the individual 

could be exposed. Initially, a PuBe neutron spectrum was used for calibration. The PuBe neutron 

spectrum has an average energy of about 4 MeV and is considerably richer in fast neutrons than the 

fission spectrum (with average energy of about 1 MeV), and so is likely to result in a calibration factor 

(i.e., tracks/cm2

-n) that would underestimate the fast neutron dose [41]. Any moderation of the fission 

spectrum encountered under field conditions would further exacerbate the underestimate.

Figure A-3. NTA film neutron energy response (IAEA 1990).

NTA film has a number of severe limitations that lead to large uncertainties in neutron dosimetry. As 

noted, the response of NTA film was highly dependent on neutron energy and angle of incidence with 

respect to the plane of the film (Cheka 1954; Lehman 1951; Kathren, Prevo, and Block 1965; Kathren 

1967). Another limitation of NTA film is its use dose range. The LLD for fast neutrons typically 

corresponds to a dose of about 100 mrem. For fast neutron fluences corresponding to a few rem, the 

track density becomes so great that accurate dosimetry becomes difficult if not impossible.

NTA films respond to other radiation qualities (photons, betas, etc.). Concomitant photon dose, which 

results in film blackening, can render track counting difficult, introduce errors and, if the optical density 

produced by the photon exposure is sufficiently great, obscure proton recoil tracks altogether. Track 

counting itself is questionable; different persons given identical sections of an exposed NTA film to 

count often produce highly variable track counts. It has been long recognized that, in the dosimetry 

laboratory, human factors associated with reading large numbers of neutron films under a microscope 

can significantly affect neutron dosimetry results. Research conducted at the Rocky Flats Plant in 

1994 reevaluated neutron doses for selected plutonium workers (ORAUT 2007). This research 

indicated that the original evaluations of films could have contained significant errors and that the 

resulting neutron doses could be significantly higher or lower than the doses actually received. The 

ATTACHMENT A

DOSIMETRY TECHNOLOGY

Page 10 of 20
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 91 of 121

degree of variation is a function of the track length and the personal style of the individual doing the 

counting. Uncertainty from this source alone could easily be as great as a factor of 2.

The latent image produced by proton recoil tracks in NTA film is highly susceptible to fading before 

development. Fading is a function of time after exposure and is particularly severe in high humidity 

(as much as 75%/wk) but can be minimized and largely controlled if the films are sealed in a moistureproof pouch before use. Humidity-induced latent image fading is probably not a significant source of 

uncertainty or dose underestimation at NTS because of the low humidity, although some latent image 

fading is likely if the wearing interval exceeds a month.

Given these limitations, neutron dose estimates made with NTA film are likely to have a high degree 

of uncertainty and generally will underestimate the actual dose from fast neutrons. As noted in Table 

6-1, this leads to the application of a bias for neutron data obtained with NTA film when detailed 

information about the neutron fields is not available (see Section 6.3.4.3).

A.2.2

The primary advantages of TLD albedo dosimeters are the high sensitivity compared to NTA film and 

the availability of automated TLD readers for rapid reading. The primary disadvantage is that the 

energy response does not match the personal dose equivalent response, so they are highly energydependent (Figure A-4). The energy response can be improved slightly by dosimeter encapsulation, 

as used in the NTS dosimeter design. However, the response is still highly dependent on the neutron 

spectrum.

TLD Albedo Detectors

Figure A-4. Neutron energy response of NTS albedo dosimeters

(IAEA 1990).

For albedo dosimeters, neutron fields can be put in four categories based on their relative spectra 

characteristics: (1) reactors, linear accelerators, and accelerators for medical therapy; (2) nuclear fuel 

fabrication areas; (3) radioactive neutron sources; and (4) high-energy accelerators with little or no 

shielding. Within a neutron spectral class, neutron response relative to HP(10) does not vary by more 

than a factor of 2. The large energy dependence is still a big disadvantage. The advantage to TLD 

ATTACHMENT A

DOSIMETRY TECHNOLOGY

Page 11 of 20
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 92 of 121

albedo detectors, compared to film and to a lesser degree CR-39, is that they detect neutrons of all 

energies and have simple automatic TLD readouts.

Calibration curves have been established for working areas that can reduce workplace-dependent 

changes of albedo response within ±30%. Depending on the neutron field, the lowest detectable dose 

using albedo TLDs varies from 5 to 20 mrem. Albedo dosimeters can be combined with TEDs for 

separate measurement of fast neutrons. In a combination detector, the albedo detector serves as the 

basic neutron detector for screening.

A.2.3

The TED is a dosimetry-grade polymer called CR-39. When fast neutrons interact with the plastic, 

submicroscopic damage trails are created. These trails can be enlarged using chemical etching 

techniques to form tracks visible under a 400-power microscope. The tracks can be made more 

visible using electrochemical etching that causes breakdown to form subsurface trees that are easily 

counted under low magnification (Griffith and Tommasino 1990). 

Track Etch Detectors

The NTS TED dosimeter consists of three CR-39 foils (i.e., individual dosimeter pieces), cut from 

0.025-in. (0.0635-cm)-thick sheets covered on both sides with 0.005-in.-thick polyethylene film. They 

are heat sealed under an opaque blister to a plasticized card. A bar code label is applied, and (for 

personnel service) the assembly is placed in a cavity in the NTS Personnel Dosimeter.

A three-step etching process is used to develop the damage tracks that result from neutron 

interactions. After removing the polyethylene film, the track detectors are first etched for 45 min at 

60°C in 6.5N KOH. This is followed by the electrochemical etching step. An alternating potential of 

3 kV is applied across the track detector in the etch bath at a frequency of 60 Hz for 3 hr. The 

resultant tracks are amplified, forming “trees” under the track detector surface. A third step using 3 kV 

but at 2,000 Hz produces tracks that are more uniform and easily recognized as tracks. The final step 

is referred to as the blow-up stage of the process. This produces well-defined round or elliptical holes 

large enough to be seen and counted with low (4-power) magnification. The tracks are recorded with 

a television camera interfaced to a commercial bacteria colony counter that can be used to count 

several standard 0.09-cm2 fields. The dose equivalent is determined by a software program that 

converts the number of foil net tracks to a dose equivalent.

The TEDs were initially calibrated with an unmoderated 252Cf source at Pacific Northwest 

Laboratories. Secondary calibration is provided by the NTS 238PuBe neutron source. Secondary 

calibration-check TED foils are processed with each batch of personnel monitoring foils to determine 

the RCF for the processing run. Primary and secondary calibration data are recorded in units of 

millirem per track per square centimeter in the TED program. The tracks from the calibration check 

foils processed with each batch of foils are compared to the primary and secondary calibration data to 

determine the dose equivalent conversion factor for the process batch. The energy response of the 

NTS track etch dosimeters is shown in Figure A-5.

Background exposure is subtracted from TED results to eliminate the TED exposure that is not part of 

the individual’s occupational dose. The natural or background neutron radiation level is extremely low 

and, therefore, NTS uses the standard practice of using control TEDs. Control TEDs are prepared 

along with batches of TEDs for issue and are retained in the low-background dosimetry operations 

ATTACHMENT A

DOSIMETRY TECHNOLOGY

Page 12 of 20
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 93 of 121

facility until the corresponding personnel TEDs are processed. Control, calibration-check, and 

personnel foils are processed together, and the number of tracks in control foils is subtracted from the 

Figure A-5. Neutron energy response of NTS TED (Allen and Schoengold 1995).

number of tracks in personnel and calibration-check foils. The calibration-check results are used to 

determine the correct dose equivalent calibration factor.

A.2.3.1 Dose Algorithms

The CR-39 TED is known to have a nonlinear response, with the observed or measured dose 

equivalent being a slight over-response in comparison with the delivered dose equivalent for low 

doses, then changing to an under-response as the dose increases. Because the nonlinearity can vary 

with CR-39 processing methods, specific data for the NTS TED system were collected. A series of 

TED exposures was made using an unmoderated PuBe source and appropriate conversion factors 

relating the source to an equivalent unmoderated 252Cf source. Two TEDs (six foils) each were 

exposed to 12 dose equivalents between 130 and 6,505 mrem. The results ranged from 126 to 

2,930 mrem, as shown in Figure A-6.

TED algorithms were developed by exposing many TEDs to calibrated neutron exposures and 

comparing processed results with calculated and known dose equivalents. Irradiations ranging from 

about 100 to 7,000 mrem were used to determine the shape of the curve relating true dose equivalent 

to observed dose. Linear regression analysis of the data pairs (true and observed dose) showed that 

a second-degree polynomial fit the relationship between true dose equivalent and indicated for 

uncorrected dose equivalent. That polynomial, shown below, is the algorithm used by the TED 

program to calculate the neutron dose equivalent HT from the TED uncorrected dose equivalent HI: 

HT = 0.77HI + 0.000459HI

2 (A-3)

HI in the above equation is calculated by the TED program from the number of net tracks and the 

millirem to track calibration factors.

ATTACHMENT A

DOSIMETRY TECHNOLOGY

Page 13 of 20

ATTACHMENT A

DOSIMETRY TECHNOLOGY

Page 14 of 20
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 94 of 121

Figure A-6. Electrochemically etched CR-39 dose response

(Allen and Schoengold 1995).

Because the Panasonic UD-802 TLD has a very slight response to neutrons, two algorithms are 

necessary to determine dose equivalent when exposure includes photons and neutrons. One 

algorithm corrects TLD response for neutron interference; the other converts TED results to dose 

equivalent. The deep dose equivalent is the sum of the deep dose from photons and the dose from 

neutrons. Therefore, in all neutron exposure cases, a TLD evaluation is performed to determine the 

deep dose from photons.

A.3 QUALITY ASSURANCE

The QA program was established and has been maintained through adherence to QC procedures and 

practices. As required by DOE rules for protecting individuals from ionizing radiation, the program 

includes internal audits at intervals of 3 yr or less. The QC personnel for the external dosimetry 

program monitor and test program operations, data records, and performance. The effectiveness of 

the QA program is demonstrated through satisfactory completion and maintenance of DOELAP 

accreditation (Allen and Schoengold 1995). 

Dosimeters are carefully tested before being put into use, and their performance is routinely 

monitored. Acceptable processing and recording equipment calibration and operation is verified by 

internal QA reviews and by participation in external assessments. For example, REECo Health 

Protection Department procedure, Thermoluminescent Dosimeter Quality Assessment (REECo 

1995c), outlines the methods for assessment of the performance and adequacy of TLD issuing, 

processing, and reporting techniques.

The NTS external dosimetry program was one of the early programs accredited by DOE. 

Accreditation was initially requested in 1989 and updated in 1990. The onsite assessment was 

conducted in 1991, and accreditation was granted in 1992. DOELAP accreditation requires that 

dosimetry operations satisfy specific standards for accuracy of measurements, records, reports, and 

QA activities.
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 95 of 121

The NTS external dosimetry program has maintained accreditation in the following DOELAP 

categories:

I. Low-energy photon (high dose)

II. High-energy photon (high dose)

IIIA. Low-energy photon

IIIB. Low-energy photon (plutonium)

IV. High-energy photon

VA. Beta

VI. Neutrons (unmoderated, 252Cf)

VII. Mixtures

III + IV

III + VA

IV + VA

III + VI

IV + VI

The accredited categories are based on possible accident scenarios and probable operational 

exposure conditions at NTS and include all DOELAP categories except two, which were omitted 

because:

• Category VB, "Beta Particles - Special" – uranium exposure environment does not exist at 

NTS. Therefore, TLDs have not been calibrated for dose measurements for beta particle 

radiation from natural or depleted uranium slabs.

• Category VI, "Neutron 252Cf (moderated)" – the unmoderated neutron category more closely 

approximates the neutron energy spectra in the NTS occupational environment.

The bias values and uncertainties in Table A-3 are based on DOELAP performance testing for TLD 

systems within 30% at an approximate 95% confidence level (Allen and Schoengold 1995).

Table A-3. Bias values and uncertainties for NTS DOELAP accreditation categories (Allen and 

Schoengold 1995). 

Category

Bias(B) and standard deviation (S) (mrem)

Deep dose Shallow dose

B S B S

I. Low-energy photons (X-ray) – high dose 0.036 0.065 N/Aa N/A

II. High-energy photons – high dose 0.046 0.070 N/A N/A

IIIa. Low-energy photons (X-ray) – general 0.001 0.117 0.03 0.097

IIIb. Low-energy photons (X-ray) – plutonium environments 0.022 0.142 0.011 0.134

IV. High-energy photons 0.099 0.048 0.039 0.058

V. Beta particles – general N/A N/A 0.091 0.036

a. N/A = not applicable.

A.4 BETA PARTICLE FIELD CHARACTERISTICS

The range of a beta particle of known energy can be rigorously calculated by integration of the –dE/dx

equation, which shows that the rate of energy loss is a complex function directly proportional to the 

ATTACHMENT A

DOSIMETRY TECHNOLOGY

Page 15 of 20
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 96 of 121

number of atoms per cubic centimeter of absorber and to the Z of the absorber, and is exponentially 

related to the kinetic energy of the beta particle (Cross et al. 1982). However, a reasonably good 

approximation of the range R of a beta particle of energy E (in megavolts-electron) in any medium can 

be calculated in terms of density thickness (in milligrams per square centimeter) from the following 

empirical equation:

R = 412E(1.265-0.0954 lnE) (A-4)

which holds over the energy range 0.01 to 2.5 MeV. For energies above 2.5 MeV, the appropriate 

empirical relationship is:

R = 530E – 106 (A-5)

A somewhat less exact but still reasonable approximation for the range of a beta particle in units of 

grams per square centimeter is to divide the beta particle energy by 2: a 3-MeV particle would have a 

range of 1.5 g/cm2

, which corresponds to 36 ft, approximately the same value (within about 3%) 

obtained using either of the above equations.

Because of attenuation of beta radiation in air, the external dose associated with a fallout field is 

strongly dependent on distance. As one moves away from the beta field, more and more of the lower 

energy particles in the spectrum reach the end of their range and no longer contribute to the dose; 

1 m of air is sufficient to attenuate all betas with energies below about 400 keV. And, as already 

noted, 36 ft of air is sufficient to absorb virtually all beta radiation associated with a fallout field, so 

beyond this distance there is zero beta dose. In contrast, as one moves closer to the fallout beta field, 

the external beta dose rate rises more rapidly than the concomitant gamma dose rate. Therefore, the 

beta-to-gamma dose rate ratio is a function of distance from the fallout field; at distances greater than 

36 ft, the ratio is zero because there is no beta dose, as noted above. Calculations by Sondhaus and 

Bond (1955) on persons highly exposed to Marshall Islands weapons fallout contamination from the 

Bravo event indicate that, relative to the average whole-body gamma dose from a fallout field, the 

corresponding beta dose would range from about a factor of 2 at the head to about a factor of 10 at 

the bottom of the feet (bare) (standing on the contamination). At 1 m above the ground, the 

beta:gamma dose rate ratio was 3. This is consistent with the data obtained onboard a contaminated 

ship that showed a beta:gamma ratio of 10 at the deck level, 8 at knee level, and about 5 at film 

badge level (about 4 ft above the deck) (Kathren 2006). 

The above results appear inconsistent with the observations of Kulp and Dick (1960), who reported 

beta dose rates at 1 in. from a contaminated aircraft surface to be typically 10-fold greater than 

gamma dose rates at 1 ft from the surface. When a correction is made for the distance by using 1/d

relationship for an infinite plane source to obtain the gamma dose rate at 1 in., the beta:gamma dose 

rate ratio falls to less than 1. However, the Kulp and Dick measurements involved vastly different 

geometries with opportunities for beta shielding and therefore might not be applicable to the type of 

exposures expected in the field at NTS. A more rigorous approach was taken by Broido and Teresi 

(1961) who evaluated the surface dose rate from beta radiation from a fallout field in relation to the 

gamma dose, and found the beta dose to be 13 times that of the gamma dose. Comparing the beta 

surface dose to the gamma dose at 1 m, Broido and Teresi observed a ratio of 40. In a study by 

Black (1962) in which actual measurements were made of doses to troops crawling through a 

contaminated fallout field, the beta:gamma dose ratio was approximately 7.

ATTACHMENT A

DOSIMETRY TECHNOLOGY

Page 16 of 20
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 97 of 121

In these studies, the concern was the dose from contamination deposited on the skin, a situation 

rather different from the external exposure situation that could be encountered in the field at NTS and 

that would likely produce a higher beta:gamma dose ratio. Although numerous theoretical and 

empirical studies have been made of gamma radiation doses above fallout-contaminated ground, 

there have been few such studies of beta radiation doses. Review of operational monitoring logs from 

several Plumbbob detonations indicate, on the basis of survey meter readings, a beta:gamma dose 

rate ratio in the range of slightly greater than unity to a maximum of about 3, regardless of distances 

above the field (ranging to 3 ft), and for times ranging from 1 to 4 wk after the detonation. 

Beta:gamma dose rate ratios did not appear dependent on dose rate but, according to a study by 

Barnaby (1957), vary with time. At 1 m above the surface of contaminated ground, Barnaby found a 

beta:gamma dose ratio of 27 at 1.5 hr after detonation, dropping rapidly to a minimum of about 3 at 10 

to 20 days after detonation, and then increasing again to about 30 at 400 days after detonation. This 

is not necessarily inconsistent with observations at NTS, which were mostly made from a few days to 

a few weeks after detonation.

Given the above discussion, for beta dose reconstruction in the absence of personnel monitoring data 

(i.e., exposures before 1967), dose reconstructors should assume that a reasonable estimate of beta 

dose favorable to claimants can be obtained by ratio with gamma dose. To this end, some provisional 

values are presented below (with the caveat that they are subject to revision as additional information 

is obtained). For beta doses incurred in a mixed beta-gamma radiation field a few days to a few 

weeks after detonation for which the gamma dose is known, a beta:gamma dose ratio of 5 would 

seem to be reasonable, while allowing a measure of overstatement to ensure favorability to claimants. 

In other words, the beta dose would be 5 times the measured gamma dose. If the mixed field 

exposure occurred less than 2 to 3 d after detonation, or more than 6 wk after detonation, a 

beta:gamma dose ratio of 10 is provisionally suggested. These provisional dose ratios take no credit 

for shielding or attenuation of beta particles by clothing (which could reduce them by a factor of 2 or 

more) and assume a 1-m distance from the fission product field with no overburden, leaching, 

vegetation, or other material or weathering action that would reduce the ratio, however likely this could 

be.

Specific information on beta:gamma ratios for specific radionuclides, including the special case of 

immersion in iodine and noble gas clouds is presented in Attachment C. 

A.5 NEUTRON FIELD CHARACTERISTICS

As indicated in Section 6.3.5.3, primary NTS operations with neutron exposure potential have been:

• Low-level waste

• Nuclear device assembly

• NRDS and BREN (Bare Reactor Experiment Nevada) tower calibrations and operation

• PLUTO reactor (nuclear-powered ramjet engine) 

• Nuclear explosive assembly using special nuclear material

• Down-hole well logging

• Neutron detection instrument calibration facilities

The neutron source spectra from these operations were either from fission of uranium and TRU 

nuclides or isotopic sources involving n,alpha reactions, such as 238PuBe or 241AmBe. Moderation or 

ATTACHMENT A

DOSIMETRY TECHNOLOGY

Page 17 of 20
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 98 of 121

scattering of the source neutrons results in a “softening” of the spectrum, with an increase in the 

fraction of low-energy neutrons present. This, in turn, results in (1) lower RBEs, quality factors, or 

radiation weighting factors, and (2) changes in the effective calibration factors for the energydependent dosimeters and instruments used for personal neutron dose assessment.

There is no indication that neutron spectral measurements were conducted at NTS for radiation 

protection purposes. However, IAEA has compiled extensive neutron spectral functions, instrument 

and dosimeter response functions, and dosimetric quantity response functions (Ing and Makra 1978; 

IAEA 1990, 2001). Because the source spectra for NTS operations are limited to those discussed 

above, adequate simulation can be obtained by selecting the proper neutron production mechanism 

and moderation and scatter conditions from the spectral catalogs that best simulate those for NTS 

operations. For example, Godiva is actually the reactor used on the BREN tower.

Figures A-7 to A-9 show the effect of scattering and moderation on the neutron spectra. Figure A-10 

and Table A-4 further illustrate the impact of spectral softening from scatter and moderation; that is, 

the fraction of HP(10,0) due to neutrons with energies less than the indicated value. For example, with 

the “hard,” unmoderated 241AmBe source spectrum, about 74% of the dose is due to neutrons above 

2 MeV. The PuBe spectrum at 1 m is probably more characteristic of an unmoderated calibration 

source spectrum, with about 63% of the HP(10) due to neutrons above 2 MeV. In contrast, consider 

the much softer 252Cf source spectrum in a 15-cm diameter D2O sphere where 9% of the dose comes 

from neutrons below 0.1 MeV and only 31% of the dose comes from neutrons with energies greater 

than 2 MeV. 

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

1.E-09 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1.E+01

Neutron energy - MeV

Fluence/unit log energy (lethargy)

Bare 252Cf

Bare 252Cf

w/room

scatter

Figure A-7. Lethargy neutron spectra for 252Cf calibration source with and without room scatter (IAEA 

2001).

ATTACHMENT A

DOSIMETRY TECHNOLOGY

Page 18 of 20
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 99 of 121

1.E-03

1.E-02

1.E-01

1.E+00

1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1.E+01

Neutron energy - meV

Fluence/unit energy (lethargy)

252Cf - ISO

252Cf in

10.16 cm

CH2

252Cf in

22.86 cm

CH2

252Cf in

22.86 cm

CH2 w/10.2

cm Fe cover

Figure A-8. Lethargy neutron spectra for 252Cf with various thickness moderators (IAEA 2001).

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1.E+01

Neutron energy - MeV

Fluence/unit energy (lethargy)

241AmBe

w/o room

scatter

241AmBe

w/room

scatter

Figure A-9. Lethargy neutron spectra for an 241AmBe calibration source with and without room 

scatter (IAEA 2001).

ATTACHMENT A

DOSIMETRY TECHNOLOGY

Page 19 of 20
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 100 of 121

0

0.2

0.4

0.6

0.8

1

1.2

1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1.E+01

Neutron energy, En - MeV

Fraction of Hp(0,10) <En

252Cf - ISO 252Cf D2O Mod. - ISO

AmBe - ISO TRU Plant, Pu Repro. Plant, Heavily shielded

Bare 238PuO2 at 100 cm Godiva Spectrum

Figure A-10. Fraction of HP(10,0) due to neutrons with energies less than En. 

Table A-4. Fraction of HP(10,0) for neutrons with energies less than En. 

Energy

En – MeV

Cf-252

ISO

Bare 

Cf-252 

w/o 

shadow 

cone

Cf-252 D2O 

Mod. - ISO

Cf-252 in 

D2O 

w/o shadow 

cone

241AmBe 

ISO

AmBe 

w/o 

shadow 

cone

PuBe 

at 1m

TRU Plant, Pu 

Repro. Plant, 

heavily 

shielded

Bare 

238PuO2

at 100 

cm

Godiva 

spectrum

1.00E-02 0.000 0.004 0.018 0.030 0.000 0.003 0.002 0.028 0.012 0.000

1.99E-02 0.000 0.005 0.027 0.039 0.000 0.003 0.002 0.032 0.013 0.000

5.01E-02 0.001 0.008 0.050 0.064 0.000 0.005 0.003 0.046 0.014 0.001

1.00E-01 0.003 0.017 0.089 0.106 0.001 0.010 0.005 0.080 0.021 0.008

1.99E-01 0.017 0.042 0.155 0.181 0.009 0.028 0.027 0.162 0.058 0.043

3.98E-01 0.072 0.114 0.228 0.283 0.037 0.070 0.066 0.321 0.225 0.158

5.01E-01 0.110 0.158 0.269 0.328 0.053 0.099 0.088 0.392 0.331 0.226

7.94E-01 0.236 0.292 0.361 0.462 0.099 0.169 0.158 0.549 0.544 0.394

1.00E+00 0.325 0.382 0.407 0.516 0.127 0.212 0.242 0.625 0.646 0.492

1.25E+00 0.430 0.486 0.476 0.579 0.158 0.260 0.310 0.694 0.744 0.591

1.99E+00 0.662 0.708 0.668 0.775 0.262 0.376 0.373 0.814 0.842 0.781

3.98E+00 0.927 0.939 0.916 0.961 0.660 0.683 0.744 0.931 0.911 0.958

5.01E+00 0.968 0.969 0.963 0.969 0.813 0.806 0.871 0.956 0.930 0.983

7.94E+00 0.999 0.988 0.998 0.992 0.995 0.969 0.997 0.989 0.977 1.000

1.00E+01 1.000 0.993 1.000 0.994 1.000 0.991 1.000 0.999 0.987 1.000

1.58E+01 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000

ATTACHMENT A

DOSIMETRY TECHNOLOGY

Page 20 of 20
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 101 of 121

TABLE OF CONTENTS 

SECTION TITLE

B.1 Introduction ..........................................................................................................................

PAGE

101

B.2 Work Area and Operation-Dependent Energy Group Photon Fractions................................ 101

B.1 INTRODUCTION

Radionuclides with photon energies greater than 250 keV contribute a significant fraction of the 

penetrating (Eγ ≥ 30 keV) photon exposures at NTS. If the fraction of photons with energies above 

250 keV in a given area and operation is known, it is possible to make a more realistic, yet favorable 

to claimants, estimate of the fraction of exposure or personal dose equivalent in this higher energy 

group.

B.2 WORK AREA AND OPERATION-DEPENDENT ENERGY GROUP PHOTON 

FRACTIONS

ORAUT (2004, Table 2-2) presents an inventory of the radionuclides encountered at NTS by area and 

operation. This table is used as a basis for estimating the likely exposure and dose equivalent 

fraction in the energy groups 30-to-250-keV and >250-keV for these areas and operations. The 

photon (gamma and X-ray) decay properties for each nuclide has been obtained from the WWW 

Table of Radioactive Isotopes found on the Internet at URL: http://ie.lbl.gov/toi/nucSearch.asp 

(Firestone and Ekström 2002). The fraction of photons per decay have been determined in each of 

the three energy groups – <30-keV, 30-to-250-keV, and >250-keV – for each nuclide (Griffith 2005). 

A lower energy cutoff of 10 keV was used for the photons below 30 keV because these are highly 

unlikely to contribute a significant fraction to the external exposure.

Using the radionuclide inventory presented in Table 2-2 and assuming equal total photon contribution 

for each nuclide in any given area and operation, the photon fraction production has also been 

estimated for each area and operation. However, two adjustments have been made to these data. 

First, it is assumed that, because of the low penetrating power of photons less than 30 keV, 50% of 

these will be attenuated in surrounding material before they can contribute to individual exposures. 

Second, 20% of the photons with energies greater than 250 keV are assumed to be scattered, and 

result in an increased contribution to the 30-to-250-keV energy group, thus reducing the contribution 

from the higher energy photons accordingly. The results are listed in Table B-1.

ATTACHMENT B

ENERGY GROUP ALLOCATION OF PHOTON EXPOSURE AND PERSONAL DOSE 

EQUIVALENT BY NTS WORK AREA AND OPERATION

Page 1 of 2
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 102 of 121

Table B-1. Work area and operation-dependent photon fractions. 

Operation Areas

Photon fraction Adjusted photon fractiona

<30 

keV 

30–250 

keV

>250 

keV

<30 

keV 

30–250 

keV

>250 

keV

Drillback operations 1–10 and 18–20 0.06 0.36 0.58 0.03 0.50 0.47

Reentry and mineback operations 1, 12, 15, and 16 0.08 0.22 0.70 0.04 0.38 0.58

Routine tunnel operations 1, 12, 15, and 16 0.00 0.03 0.97 0.00 0.22 0.78

Decontamination facility 6 0.24 0.28 0.48 0.14 0.43 0.44

Treatability test facility 25 0.22 0.32 0.47 0.12 0.46 0.42

Atmospheric safety test areas 5 and 11 None indicated None indicated

Atmospheric weapons test areas 1–5, 7–11, and 18 0.18 0.31 0.51 0.10 0.45 0.45

Low-level waste site 3 0.47 0.25 0.28 0.31 0.40 0.29

Low-level waste site 5 0.34 0.31 0.36 0.20 0.45 0.34

Radiation instrument calibration 6 and 23 0.54 0.25 0.22 0.36 0.40 0.24

Radiograph operations 23 0.01 0.04 0.96 0.00 0.23 0.76

Well logging operations 1–10 and 18–20 0.49 0.23 0.29 0.32 0.38 0.30

Nuclear explosive/device assembly 6 and 27 0.73 0.27 0.00 0.57 0.43 0.00

Nuclear rocket development 25 and 26 0.04 0.29 0.67 0.02 0.43 0.55

Radioactive source storage 6 and 23 0.40 0.25 0.32 0.26 0.41 0.33

Radiochemistry and counting 

laboratories

6 and 23 0.16 0.25 0.59 0.09 0.40 0.51

a. Adjustment factors: 

<30-keV contribution = 50 % of relative photon contribution to account for attenuation of low-energy photons.

30–250 KeV = photon contribution in that energy range + 20% of >250-keV contribution to account for scatter of high-energy photons.

>250 keV = 0.8 × high-energy photon contribution to account for loss due to scatter.

ATTACHMENT B

ENERGY GROUP ALLOCATION OF PHOTON EXPOSURE AND PERSONAL DOSE 

EQUIVALENT BY NTS WORK AREA AND OPERATION

Page 2 of 2
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 103 of 121

TABLE OF CONTENTS 

SECTION TITLE

C.1 Introduction ..........................................................................................................................

PAGE

103

C.2 Beta-Photon Ratios from Standing on a Contaminated Surface Following 

Nuclear Test or Reactor Operation ...................................................................................... 104

C.3 Beta-Photon Ratios from Immersion in a Contaminated Cloud Following 

Nuclear Test or Reactor Operation....................................................................................... 107

C.4 Beta-Photon Ratios for Other Geometries............................................................................ 107

C.4.1 Photon Dose Constants............................................................................................107

C.4.2 Beta Dose Rates ......................................................................................................113

C.5 Summary of Beta-Photon Ratios.......................................................................................... 114

C.1 INTRODUCTION

As beta particles from sources outside the body enter tissue, the dose falls off rapidly with depth, and 

tissues and organs lying deeper than 10 mm in the body are unaffected. Thus, beta particles are 

appropriately ignored in considering external dose to most tissues and organs, which lie deeper than 

10 mm; for them the appropriate quantities are gamma dose and neutron dose. The two exceptions 

are the skin, with its sensitive component (basal cells) at a depth of 0.07 mm, and the eye, with its 

sensitive component (lens) at a depth of 3 mm. The potential contribution from beta particles should 

be considered whenever the dose to skin or the lens of the eye is assessed.

As a result of a range of operations at NTS, a range of radionuclides can be found across the Site. 

An inventory of these nuclides is found in Table 2-2. However, nuclear testing and reactor operations, 

including nuclear rocket and ramjet tests, result in a much larger inventory in areas where these 

operations are conducted. The situation is made much more complex by the time dependence of the 

radionuclide inventory following detonation or reactor operation. This time dependence results in a 

variation in the radiological properties of the inventory that also varies with time. An extensive set of 

publications by Harry Hicks at LLNL (Hicks 1981a to 1981i, 1982, 1984) has produced a vast set of 

data covering 177 fission and activation products for periods from 1 hr to 50 yr following detonation or 

reactor operation. Surface roughness effects are simulated by using values of (mR/h)/(μCi/m2

) for a 

relaxation length of 0.16 g/cm2 (Beck 1980). According to Beck, the concentration of fallout varies 

exponentially with soil depth, Z, according to the relation C = Coe-αZ

, and he defines relaxation length 

as 1/α. The Hicks publications represent a very detailed source term for beta-photon ratio 

calculations.

ATTACHMENT C

BETA-PHOTON RATIO ESTIMATES FOR NTS WORK AREAS AND OPERATIONS

Page 1 of 14
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 104 of 121

C.2 BETA-PHOTON RATIOS FROM STANDING ON A CONTAMINATED SURFACE

FOLLOWING NUCLEAR TEST OR REACTOR OPERATION 

The fundamentals of the method for estimating external dose to the skin and eye from standing on 

contaminated ground are summarized as follows:

1. Beta dose to the skin or lens of the eye from external sources is accrued with the gamma dose 

from radioactive fallout, contamination, or neutron-induced radionuclides. As a result, the beta 

dose is proportional to the gamma dose, and its relative magnitude can be expressed by a 

beta-to-gamma dose ratio.

2. The beta-to-gamma dose ratio depends on radionuclide decay and distribution and on 

geometric relationships between the exposed individual and the radiation source. Gamma and 

beta energy spectra are interdependent functions of time. Consequently, the beta-to-gamma 

dose ratio depends on time since detonation.

3. Because of the attenuation characteristics of electrons in matter, beta dose assessments 

depend more critically than gamma dose assessments on geometry and the shielding material 

between the radioactive source and the exposed individual. Consequently, the nature of 

specific job- or task-related activities and their associated protective measures entails special 

attention and evaluation in determining a beta dose component.

4. Beta doses to skin are evaluated at the anatomic location where a skin cancer has been 

diagnosed. The depth for evaluation is 0.07 mm, the conventional depth of the basal-cell layer 

of the skin, which is assumed to be the tissue at risk for skin cancer. Beta doses to the lens of 

the eye are assessed at a depth of 3 mm below the front surface of the eye, where the tissue 

at risk for posterior subcapsular cataract development is assumed to be.

5. A beta energy greater than 0.07 MeV is required to penetrate the dead epidermal layer, so 

beta particles with energies less than that are not included in dose assessments.

6. Skin and eye doses are assessed as the sum of the applicable “high-sided” beta and “highsided” gamma doses (neutron doses presumably are included if they are significant). 

7. Fallout deposited on a surface is considered to be a semi-infinite plane isotropic source, and 

decontamination activities are considered in evaluating beta doses. Assessments of skin 

doses are simplified by ignoring attenuation of electrons by large fallout particles that contain 

volume-distributed activity, particle and photon scattering due to surface roughness, particle 

and photon attenuation due to penetration into a radioactive surface, and radioactive-source 

depletion due to weathering, chemical dissociation, or environmental transport (concentration 

or dispersion). Those simplifications have the effect of making the calculated doses 

overestimates.

Barss (2000) provides methods and tables useful for assessing beta dose. They include separate 

tables of beta-to-gamma dose ratios from exposure to fission products on the ground as a function of 

time since detonation for NTS and Pacific tests (in the Pacific, one table applies to Operation 

CASTLE, Test BRAVO fallout, and a second table applies to all other tests). A separate table of betaATTACHMENT C

BETA-PHOTON RATIO ESTIMATES FOR NTS WORK AREAS AND OPERATIONS

Page 2 of 14
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 105 of 121

to-gamma dose ratios is provided for activation products in soil. The beta-to-gamma dose ratios are 

calculated from published beta-particle and gamma-ray spectra. For illustration, the beta-to-gamma 

dose ratios from the table for Pacific tests are plotted in Figure C-1. The figure illustrates the 

substantial variation of beta-to-gamma dose ratios with time after detonation and height above 

ground. The beta-to-gamma dose ratios for the lens of the eye are much smaller because the greater 

depth of the sensitive tissue (3 mm for lens vs. 0.07 mm for skin) leads to much more attenuation.

Figure C-1. Beta-to-gamma dose ratios for contaminated surfaces used 

at Pacific tests (Barss 2000).

Activation products are distributed with depth in soil because they originate primarily by interactions 

with neutrons that penetrate into the soil rather than in deposition from the atmosphere. Thus, betato-gamma dose ratios of activation products are small because most of the activation products are 

deeper in the soil than the range of the emitted beta particles.

Extensive calculations of beta and photon dose conversion factors for skin from (1) surface 

contamination and (2) air immersion are now available for 825 radionuclides (Eckerman et al. 1999),

based on the work of Kocher and Eckerman (1981). The geometry involves the use of a reference 

adult phantom placed (1) on a contaminated ground plane and (2) immersed in a semi-infinite cloud of 

contaminated air. The CD, FGR_13, does not provide separate conversion factors for beta and 

photons. However, they are available for skin only from Eckerman (2006b).

The Hicks report presents the surface contamination following each event in three tables of 

radionuclide concentration in terms of μCi/m2

: zero to 21 hr after detonation (177 radionuclides); zero 

to 300 d after detonation (128 radionuclides); and zero to 50 yr after detonation (37 radionuclides). 

These data have been used together with the beta and photon dose conversion factors (Eckerman 

2006b) to calculate radionuclide inventory-weighted beta-photon ratios. Examples of the resulting 

calculations for selected atmospheric tests are shown in Figure C-2. These are quite consistent with 

the values shown in Figure C-1 for a height of about 120 cm. Both figures show a significant increase 

at later times, 100–200 d. The maximum time for Figure C-1 is 2 yr, so the decrease after that in 

Figure C-2 out to 50 yr is not shown. Figure C-3 shows the results of similar calculations for the 

effluents from nuclear rocket and ramjet tests. Tabulated beta-photon ratios for the fallout from 35 

ATTACHMENT C

BETA-PHOTON RATIO ESTIMATES FOR NTS WORK AREAS AND OPERATIONS

Page 3 of 14
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 106 of 121

1

10

100

1.E-03 1.E-02 1.E-01 1.E+00 1.E+01 1.E+02 1.E+03 1.E+04 1.E+05

Elapsed Time - Days

Beta/Photon Sv/Sv

Annie Nancy Ruth Dixie Ray Badger

Simon Encore Harry Grable Climax

Figure C-2. Calculated beta-photon ratios (Sv/Sv) for skin for 11 atmospheric tests from 

Operation Upshot-Knothole. 

1

10

100

1.E-03 1.E-02 1.E-01 1.E+00 1.E+01 1.E+02 1.E+03 1.E+04 1.E+05

Elapsed Time - Days

Beta/Photon Sv/Sv

Figure C-3. Calculated beta-photon ratios (Sv/Sv) for skin for nuclear rockets and ramjet test 

effluents. 

ATTACHMENT C

BETA-PHOTON RATIO ESTIMATES FOR NTS WORK AREAS AND OPERATIONS

Page 4 of 14
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 107 of 121

tests are listed in Table C-1. The maximum values shown in the last row of this table can serve as a 

guide for establishing default beta-photon ratios for this exposure geometry. These values can be 

modified with appropriate factors for shielding and distance (Barss and Weitz 2006).

C.3 BETA-PHOTON RATIOS FROM IMMERSION IN A CONTAMINATED CLOUD 

FOLLOWING NUCLEAR TEST OR REACTOR OPERATION

The special situation exists in which a worker might have been exposed to photons and beta particles 

by immersion in a contaminated cloud. This situation is known to have occurred due to release of 

radioiodines and noble gases during post-test drilling operations. Because of the exposure geometry, 

beta gamma ratios under such circumstances are likely to have been different than those addressed 

above.

The beta dose rate factors due to immersion in semi-infinite contaminated clouds of radionuclides in 

air have been estimated by Kocher and Eckerman (1981) and implemented in Eckerman et al. (1999). 

As for the surface contamination geometry, the separate unpublished beta and photon dose 

conversion factors for each radionuclide are available (Eckerman 2006b). Table C-2 lists the timedependent beta-photon ratios for the radionuclide inventories provided in the Hicks reports, but in a 

semi-infinite cloud. The ratios are lower than those from a contaminated surface because the betas 

are absorbed by the cloud to a greater extent than the photons. The exposure of post-test drilling 

crews to clouds of fission product noble gases (krypton and xenon), and halogens (bromine and 

iodine) has been noted. A separate calculation has been made for cloud immersion beta-photon 

ratios following 11 tests in the Range and Buster-Jangle test series using only the isotopes of those 

elements. There is very little variation in the results from one test to another, because there is little 

variation and fractionation for those nuclides. The maximum ratios are shown in Figure C-4. The 

ratios are consistently below 1.0 until the 10.7-yr 85Kr begins to dominate after about 100 d.

C.4 BETA-PHOTON RATIOS FOR OTHER GEOMETRIES

In addition to activation and fission products from nuclear tests and reactor operation, radionuclides 

have been used or handled in various areas throughout NTS. An inventory of these radionuclides is 

presented in ORAUT (2004, Table 2-2). For the most part, the geometries for exposure to these 

radionuclides is difficult to specify with any accuracy. The beta-photon ratios are also difficult to 

specify with any certainty. In many cases, such as the use of encapsulated radioactive sources, 

although the radionuclides involved might have an inherently high beta-photon ratio, the 

encapsulation is likely to absorb the associated beta particles. For the purpose of estimating betaphoto ratios for a geometry other than extended contaminated surfaces or cloud immersion, it is 

necessary to select an arbitrary, but realistic, geometry as a basis for the ratio estimates. The 

calculations presented in this attachment are based on a distance of 1 m for a point source.

C.4.1

The values of the specific gamma ray dose constant, Γ, in mSv/hr/MBq at 1 m were established for 

each radionuclide in the inventory listed in Table 2-2. The values of Γ for several of those nuclides 

are presented in Table 6.2.2 of the Handbook of Health Physics and Radiological Health (Shleien, 

Slaback, and Birky 1998). Those values are based on the work of Unger and Trubey (1981), and 

appear in Table C-3. 

Photon Dose Constants

ATTACHMENT C

BETA-PHOTON RATIO ESTIMATES FOR NTS WORK AREAS AND OPERATIONS

Page 6 of 7 

ATTACHMENT C

BETA-PHOTON RATIO ESTIMATES FOR NTS WORK AREAS AND OPERATIONS

Page 5 of 14
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 108 of 120

Table C-1. Calculated beta-photon ratios for mixed fission and activation products from Sedan and tests from Operations Ranger, 

Buster-Jangle, Tumbler-Snapper, and Upshot-Knothole—contaminated surfaces. 

Event Year

Beta-photon ratio

Hours Days Years

Overall 

averagea

1-day 

reentrya

Onsite 

during 

yeara 0. 1.0 2. 12. 1.0 10. 50. 200. 1.0 2.0 5.0 10. 20. 50.

Ranger Able 1951 10.4 7.5 6.2 8.1 6.0 2.8 5.0 7.2 15.6 47.0 18.0 13.2 14.9 15.6 12.7 8.0 24.3

Ranger Baker 9.6 7.1 5.9 7.4 5.7 2.7 4.3 7.8 19.2 43.9 17.0 8.8 9.6 10.1 11.4 7.4 21.6

Ranger Easy 12.5 8.3 6.4 6.8 4.8 2.3 4.3 7.8 19.2 43.9 17.0 8.8 9.6 10.0 11.6 7.1 21.6

Ranger Baker 2 9.3 7.0 5.9 7.5 5.7 2.7 4.3 7.8 19.2 43.9 17.0 8.8 9.6 10.1 11.3 7.4 21.6

Ranger Fox 8.2 6.7 5.9 8.3 6.5 3.0 4.8 7.3 17.0 45.9 18.0 11.9 13.4 14.0 12.2 8.0 23.5

Buster Baker 1951 5.8 5.8 5.4 7.8 6.6 3.0 3.7 7.3 18.5 41.5 16.1 6.9 7.3 7.7 10.2 7.4 18.9

Buster Charlie 8.2 6.6 5.9 8.4 6.6 3.0 4.5 6.4 14.8 45.0 17.7 12.3 13.9 14.5 12.0 8.1 21.9

Buster Dog 8.8 6.9 5.9 8.1 6.3 2.8 4.5 6.5 15.4 44.4 17.5 11.4 12.9 13.4 11.8 7.9 21.9

Buster Easy 7.0 6.3 5.7 8.6 6.9 3.1 4.4 6.5 15.2 44.5 17.6 11.6 13.1 13.7 11.7 8.2 21.5

Jangle Sugar 1951 7.2 6.1 5.6 6.6 5.2 2.8 6.5 12.3 22.8 45.7 21.9 16.9 16.5 15.8 13.7 6.5 32.9

Jangle Uncle 7.2 6.1 5.6 6.6 5.2 2.8 6.5 12.3 22.8 45.8 21.9 16.9 16.5 15.8 13.7 6.5 32.9

Tumbler-Snapper Able 1952 11.6 7.9 6.4 7.8 5.6 2.7 5.5 8.3 18.7 59.0 24.4 16.4 16.4 15.8 14.8 7.8 26.9

Tumbler-Snapper Baker 11.5 7.8 6.3 7.9 5.7 2.7 5.5 8.3 18.7 59.0 24.4 16.4 16.3 15.8 14.7 7.9 26.9

Tum.-Snapper Charlie 9.0 6.9 6.0 8.2 6.3 2.9 5.2 8.6 20.3 55.4 23.3 14.1 13.9 13.5 13.8 8.0 25.8

Tum.-Snapper Dog 7.6 6.4 5.7 8.4 6.7 3.1 5.2 8.6 20.3 55.4 23.3 14.1 13.9 13.5 13.7 8.1 25.8

Tum.-Snapper Easy 6.0 5.8 5.4 7.0 5.8 3.0 5.9 11.6 24.1 47.3 21.8 14.4 14.0 13.4 13.3 6.8 30.0

Tum.-Snapper Fox 7.1 6.1 5.6 6.8 5.4 2.9 5.9 11.6 24.1 47.5 21.9 14.5 14.1 13.5 13.4 6.7 30.0

Tum.-Snapper George 7.0 6.2 5.6 6.8 5.4 2.9 5.9 11.6 24.1 47.5 21.9 14.5 14.1 13.5 13.4 6.7 30.0

Tum.-Snapper How 5.6 5.7 5.4 7.1 5.9 3.1 5.9 11.6 24.0 47.3 21.8 14.5 14.1 13.5 13.3 6.8 30.0

Upshot-Knot. Annie 1953 6.9 6.1 5.3 6.9 5.5 2.9 5.8 11.5 23.5 45.1 20.4 13.8 13.8 13.5 12.9 6.7 29.6

Upshot-Knot. Nancy 6.9 6.0 5.5 6.9 5.5 2.9 5.9 11.6 23.8 44.1 20.5 14.0 13.9 13.5 12.9 6.7 30.0

Upshot-Knot. Ruth 4.7 5.3 5.3 6.8 5.1 2.7 6.3 11.0 21.1 49.0 22.6 16.9 16.5 15.8 13.5 6.5 31.4

Upshot-Knot. Dixie 7.6 6.6 5.6 8.4 6.6 3.0 4.4 6.6 15.7 50.8 23.0 14.2 13.9 13.5 12.9 8.1 21.5

Upshot-Knot. Ray 8.5 6.6 5.8 6.8 5.2 2.8 6.3 11.0 21.6 48.9 22.5 16.9 16.5 15.8 13.9 6.8 31.4

Upshot-Knot. Badger 8.0 6.4 5.7 6.6 5.2 2.7 5.8 11.7 24.2 43.0 19.8 13.2 13.1 12.9 12.7 6.6 29.6

Upshot-Knot. Simon 6.7 6.0 5.6 7.1 5.7 2.9 6.2 11.3 22.4 46.9 22.1 16.1 15.7 15.1 13.6 6.9 31.2

Upshot-Knot. Encore 8.3 6.7 5.9 8.6 6.8 3.1 5.4 8.5 22.4 46.9 22.1 16.1 15.7 15.1 13.7 8.3 26.6

Upshot-Knot. Harry 7.1 6.1 5.6 6.9 5.5 2.9 6.0 11.4 23.3 46.4 20.8 14.7 14.5 14.0 13.2 6.8 30.4

Upshot-Knot. Grable 6.2 6.0 5.6 8.7 7.0 3.5 5.5 8.0 18.0 57.7 25.1 17.1 16.7 16.0 14.4 8.2 26.8

Upshot-Knot. Climax 6.7 6.3 5.7 8.9 7.2 3.3 4.6 6.4 14.9 54.6 24.1 15.8 15.7 15.1 13.5 8.4 22.3

Sedan 1962 5.3 5.4 4.8 4.0 3.3 2.5 2.6 2.9 4.6 11.6 10.1 5.3 6.5 7.8 5.5 4.2 12.3

Maximum 12.5 8.3 6.4 8.9 7.2 3.5 6.5 12.3 24.2 59.0 25.1 17.1 16.7 16.0 16.0 8.8 32.9

a. The average values may be used when a reasonable approach is required or when the employee is not directly identified with an event.

ATTACHMENT C

BETA-PHOTON RATIO ESTIMATES FOR NTS WORK AREAS AND OPERATIONS

Page 6 of 14
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 109 of 120

Table C-2. Calculated beta-photon ratios for mixed fission and activation products from Operations Ranger, Buster-Jangle, TumblerSnapper, and Upshot-Knothole tests—immersion in a contaminated semi-infinite cloud. 

Event Year

Beta-photon ratio

Hours Days Years

Overall 

averagea

1-day 

reentrya

Onsite 

during 

yeara 0. 1.0 2. 12. 1.0 10. 50. 200. 1.0 2.0 5.0 10. 20. 50.

Ranger Able 1951 1.4 1.0 0.8 1.2 1.0 0.5 0.7 1.0 2.2 6.6 2.6 2.0 2.3 2.4 1.8 1.2 3.4

Ranger Baker 1.3 0.9 0.8 1.1 0.9 0.5 0.6 1.1 2.7 6.2 2.4 1.3 1.5 1.6 1.6 1.1 3.0

Ranger Easy 1.7 1.1 0.8 1.1 0.9 0.5 0.6 1.1 2.7 6.2 2.4 1.3 1.5 1.6 1.7 1.1 3.0

Ranger Baker 2 1.2 0.9 0.8 1.1 0.9 0.5 0.6 1.1 2.7 6.2 2.4 1.3 1.5 1.6 1.6 1.1 3.0

Ranger Fox 1.1 0.9 0.8 1.2 1.0 0.5 0.7 1.0 2.4 6.5 2.6 1.8 2.0 2.1 1.8 1.2 3.4

Buster Baker 1951 0.8 0.8 0.7 1.0 0.9 0.5 0.5 1.0 2.6 5.9 2.3 1.0 1.2 1.2 1.5 1.0 2.6

Buster Charlie 1.1 0.9 0.8 1.2 1.0 0.5 0.6 0.9 2.1 6.4 2.5 1.8 2.1 2.2 1.7 1.2 2.9

Buster Dog 0.8 0.8 0.7 1.0 1.0 0.5 0.6 0.9 2.2 6.3 2.5 1.7 2.0 2.1 1.7 1.0 2.9

Buster Easy 0.9 0.8 0.7 1.2 1.0 0.5 0.6 0.9 2.1 6.3 2.5 1.7 2.0 2.1 1.7 1.1 2.9

Jangle Sugar 1951 0.9 0.8 0.7 0.9 0.8 0.5 0.8 1.6 3.2 6.6 3.2 2.5 2.5 2.4 2.0 0.9 4.1

Jangle Uncle 0.9 0.8 0.7 0.9 0.8 0.5 0.8 1.6 3.2 6.6 3.2 2.6 2.5 2.4 2.0 0.9 4.1

Tumbler-Snapper Able 1952 1.6 1.0 0.8 1.2 1.0 0.6 0.8 1.1 2.6 8.4 3.6 2.5 2.5 2.4 2.2 1.2 3.9

Tumbler-Snapper Baker 1.6 1.0 0.8 1.2 1.0 0.6 0.8 1.1 2.6 8.4 3.6 2.5 2.5 2.4 2.2 1.2 3.9

Tum.-Snapper Charlie 1.2 0.9 0.8 1.2 1.0 0.5 0.7 1.2 2.8 7.9 3.4 2.1 2.1 2.1 2.0 1.2 3.5

Tum.-Snapper Dog 1.0 0.8 0.7 1.2 1.0 0.5 0.7 1.2 2.8 7.9 3.4 2.1 2.1 2.1 2.0 1.1 3.5

Tum.-Snapper Easy 0.8 0.8 0.7 1.0 0.8 0.5 0.8 1.6 3.4 6.8 3.2 2.2 2.2 2.1 1.9 1.0 4.1

Tum.-Snapper Fox 0.9 0.8 0.7 1.0 0.8 0.5 0.8 1.6 3.4 6.8 3.2 2.2 2.2 2.1 1.9 1.0 4.1

Tum.-Snapper George 0.9 0.8 0.7 1.0 0.8 0.5 0.8 1.6 3.4 6.8 3.2 2.2 2.2 2.1 1.9 1.0 4.1

Tum.-Snapper How 0.7 0.7 0.7 1.0 0.8 0.5 0.8 1.6 3.4 6.8 3.2 2.2 2.2 2.1 1.9 1.0 4.1

Upshot-Knot. Annie 1953 0.9 0.8 0.7 1.0 0.8 0.5 0.8 1.6 3.3 6.5 3.0 2.1 2.1 2.1 1.9 1.0 4.1

Upshot-Knot. Nancy 0.9 0.8 0.7 1.0 0.8 0.5 0.8 1.6 3.3 6.1 2.8 2.0 2.1 2.1 1.8 1.0 4.1

Upshot-Knot. Ruth 0.7 0.7 0.7 1.0 0.8 0.5 0.8 1.5 3.6 7.6 3.6 2.8 2.8 2.6 2.1 1.0 4.0

Upshot-Knot. Dixie 1.0 0.9 0.7 1.2 1.0 0.5 0.6 0.9 2.2 7.1 3.2 2.1 2.1 2.1 1.8 1.1 2.9

Upshot-Knot. Ray 1.1 0.9 0.8 1.0 0.8 0.5 0.8 1.5 3.0 7.0 3.3 2.6 2.5 2.4 2.0 1.0 4.0

Upshot-Knot. Badger 1.0 0.8 0.7 1.0 0.8 0.5 0.8 1.6 3.4 6.0 2.7 1.9 2.0 2.0 1.8 1.0 4.1

Upshot-Knot. Simon 0.9 0.8 0.7 1.0 0.8 0.5 0.8 1.5 3.1 6.5 3.1 2.4 2.4 2.3 1.9 1.0 4.0

Upshot-Knot. Encore 1.1 0.9 0.8 1.2 1.0 0.5 0.7 1.2 3.1 6.5 3.1 2.4 2.4 2.3 1.9 1.2 3.5

Upshot-Knot. Harry 0.9 0.8 0.7 1.0 0.8 0.5 0.8 1.6 3.3 6.7 3.0 2.2 2.2 2.1 1.9 1.0 4.1

Upshot-Knot. Grable 0.8 0.8 0.7 1.2 1.0 0.5 0.8 1.1 2.5 8.3 3.7 2.6 2.5 2.4 2.1 1.1 3.9

Upshot-Knot. Climax 0.9 0.8 0.7 1.2 1.0 0.5 0.6 0.9 2.1 7.8 3.5 2.4 2.4 2.3 1.9 1.1 2.9

Maximum 1.7 1.1 0.8 1.2 1.0 0.6 0.8 1.6 3.6 8.4 3.7 2.8 2.8 2.6 2.3 1.2 4.1

a. The average values may be used when a reasonable approach is required or when the employee is not directly identified with an event.

ATTACHMENT C

BETA-PHOTON RATIO ESTIMATES FOR NTS WORK AREAS AND OPERATIONS

Page 7 of 14
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 110 of 121

0.1

1

10

100

0.001 0.01 0.1 1 10 100 1000

Elapsed Time - Days

Beta-Photon Ratio

Figure C-4. Calculated beta-photon ratios for immersion in a contaminated 

semi-infinite cloud—maximum of values from 11 atmospheric test 

radionuclide inventories. 

Table C-3. Photon dose constants, beta dose rates, and betaphoton ratios for NTS radionuclides. 

Radionuclide 

of concern

Photon dose constant, 

Γa mSv/hr per MBq

Beta dose rate

at 1 mb D(0.07)

mGy/hr per MBq

Beta/photon

Gy/Sv

Ac-227 2.36E-06 0.0E+00 0.0E+00

Am-241 8.48E-05 0.0E+00 0.0E+00

Am-243 8.46E-05 0.0E+00 0.0E+00

AmBe-241 8.48E-05 N/A N/A

Ba-133 1.23E-04 0.0E+00 0.0E+00

Ba-140 4.45E-05 1.2E-02 2.7E+02

Cd-109 4.98E-05 0.0E+00 0.0E+00

Ce-139 5.55E-05 0.0E+00 0.0E+00

Ce-141 1.98E-05 3.1E-04 1.5E+01

Ce-143 6.89E-05 3.9E-03 5.6E+01

Ce-144 6.30E-06 D(0.07) = 9.2E-03 1.5E+03

D(10) = 1.5E-05 2.4E+00

Cf-252 1.13E-05 0.0E+00 0.0E+00

Cm-244 1.74E-05 0.0E+00 0.0E+00

Co-57 4.09E-05 4.0E-03 9.8E+01

Co-60 3.70E-04 0.0E+00 0.0E+00

Cs-134 2.70E-04 1.4E-03 5.2E+00

Cs-137c 1.03E-04 2.1E-03 2.1E+01

Eu-152 2.01E-04 3.0E-03 1.5E+01

Eu-154 2.04E-04 2.8E-03 1.4E+01

Eu-155 1.80E-05 0.0E+00 0.0E+00

Fe-59 1.79E-04 5.0E-05 2.8E-01

H-3 0.00E+00 0.0E+00 N/A

Hg-203 6.84E-05 0.0E+00 0.0E+00

I-131 7.65E-05 1.2E-03 1.6E+01

ATTACHMENT C

BETA-PHOTON RATIO ESTIMATES FOR NTS WORK AREAS AND OPERATIONS 

Page 8 of 14
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 111 of 121

Radionuclide 

of concern

Photon dose constant, 

Γa mSv/hr per MBq

Beta dose rate

at 1 mb D(0.07)

mGy/hr per MBq

Beta/photon

Gy/Sv

I-132 3.86E-04 6.9E-03 1.8E+01

I-133 1.11E-04 6.8E-03 6.1E+01

I-135 2.33E-04 5.7E-03 2.4E+01

Ir-192 1.60E-04 1.2E-03 7.5E+00

Kr-85 4.23E-07 3.5E-03 8.3E+03

Lu-174 3.06E-05 5.6E-03 1.8E+02

Mn-54 1.38E-04 2.8E-03 2.0E+01

Mo-99 3.05E-05 6.3E-03 2.1E+02

Na-22 3.62E-04 8.5E-04 2.3E+00

Na-24 5.24E-04 8.9E-03 1.7E+01

Ni-63 0.00E+00 0.0E+00 N/A

Np-237 1.25E-04 0.0E+00 0.0E+00

Pm-147 7.23E-10 0.0E+00 0.0E+00

Pu-238 2.14E-05 0.0E+00 0.0E+00

Pu-239 8.15E-06 0.0E+00 0.0E+00

Pu-240 2.03E-05 0.0E+00 0.0E+00

Pu-241 0.00E+00 0.0E+00 0.0E+00

Pu-242 1.68E-05 0.0E+00 0.0E+00

PuBe-238 N/A N/A N/A

Ra-226 3.27E-06

RaBe-226 N/A N/A N/A

Rh-100 4.55E-04 4.5E-03 9.9E+00

Rh-101 1.32E-04 2.6E-04 2.0E+00

Rh-102m 4.37E-04 1.1E-03 2.6E+00

Rh-103m 6.91E-06 0.0E+00 0.0E+00

Rh-106 3.73E-05 D(0.07) = 8.4E-03 2.2E+02

D(10) = 4.7E-05 1.3E+00

Ru-103 8.97E-05 1.9E-04 2.1E+00

Ru-106c 3.73E-05 D(0.07) = 8.4E-03 2.2E+02

D(10) = 4.7E-05 1.3E+00

Sb-124 2.88E-04 4.1E-03 1.4E+01

Sb-125 1.03E-04 6.0E-04 5.8E+00

Sn-113 4.84E-05 3.6E-03 7.4E+01

Sr-85 2.05E-04 3.2E-03 1.6E+01

Sr-89 2.21E-08 8.3E-03 3.8E+05

Sr-90/Y-90 0.00E+00 9.7E-03 N/A

Ta-182 2.09E-04 6.6E-04 3.2E+00

Tc-99m 3.32E-05 0.0E+00 0.0E+00

Te-132 7.55E-05 0.0E+00 0.0E+00

Th-228 2.14E-05 0.0E+00 0.0E+00

Th-230 1.86E-05 0.0E+00 0.0E+00

Th-232 1.85E-05 0.0E+00 0.0E+00

Th-234 2.04E-05 0.0E+00 0.0E+00

U-233 7.87E-06 0.0E+00 0.0E+00

U-234 2.10E-05 0.0E+00 0.0E+00

U-235 9.16E-05 0.0E+00 0.0E+00

U-235 Chainc 2.39E-04 1.8E-07 7.5E-04

U-238 1.76E-05 0.0E+00 0.0E+00

U-238 Chaind 6.24E-05 8.5E-03 1.4E+02

Xe-133 2.78E-05 0.0E+00 0.0E+00

ATTACHMENT C

BETA-PHOTON RATIO ESTIMATES FOR NTS WORK AREAS AND OPERATIONS

Page 9 of 14
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 112 of 121

Xe-133m 3.03E-05 0.0E+00 0.0E+00

Xe-135 5.12E-05 4.8E-03 9.4E+01

Y-88 4.82E-04 0.0E+00 0.0E+00

Y-91 5.40E-07 8.3E-03 1.5E+04

Yb-169 8.84E-05 2.6E-03 2.9E+01

Zr-95e 1.26E-4

+ 

1.30E-4

6.3E-05 2.5E-01

Zr-97e 2.92E-5 

+ 

1.26E-4

1.2E-02 7.7E+01

a. Values of the photon dose constant, Γ, that were calculated using Equation C-1 are listed 

in bold.

b. Values of beta dose rate that were calculated using the approximation in Equation C-4 

are listed in bold.

c. In equilibrium with Th-231. mSv/hr per MBq U = 235.

d. In equilibrium with Th-234, Pa-234m, and U-234. mSv/hr per MBq U = 238.

e. In equilibrium with daughter.

For Table 2-2 nuclides not presented in the Handbook of Health Physics and Radiological Health, 

Table 6.2.2, the values of Γ were calculated using the procedure of Unger and Trubey, as follows:

1. The photon (gamma and X-ray) decay properties for each nuclide has been obtained from the 

WWW Table of Radioactive Isotopes found on the internet at URL: 

http://ie.lbl.gov/toi/nucSearch.asp (Firestone and Ekström 2002).

2. The values of the specific gamma ray dose constant, Γ, were calculated using the following 

equation:

Γ = 107 [1/(4 π R2

)] Σ Si D(Ei) mSv hr-1 per MBq, (C-1) 

where R = 100 cm

Si = emission probability of each photon

Ei = energy of each photon (MeV)

D(Ei

) = dose rate per unit flux density

3. The dose rate per unit flux density is determined was follows:

Ln D(E) = A + B (Ln E) + C (Ln E)2 + F (Ln E)3

 (C-2)

where

Photon energy

MeV A B C F

0.01 to 0.03 -20.477 -1.7454

0.03 to 0.5 -13.626 -0.57117 -1.0954 -0.24897

0.5 to 5.0 -13.133 0.72008 -0.033603

4. Values of Γ for nuclides not presented in Table 6.2.2 of Shleien, Slaback, and Birky (1998) 

were calculated as above and are listed in Table C-3 in bold.

ATTACHMENT C

BETA-PHOTON RATIO ESTIMATES FOR NTS WORK AREAS AND OPERATIONS

Page 10 of 14
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 113 of 121

C.4.2

The beta dose rates as determined in this section are considered to be those at 0.07 mm. The beta 

decay data for the nuclides listed in Table C-3 were obtained from WWW Table of Radioactive 

Isotopes found on the Internet at URL: http://ie.lbl.gov/toi/nucSearch.asp (Firestone and Ekström 

2002). For comparison, beta dose rates in mGy/hr/MBq were estimated at 1 m. The range of a 350-

keV electron in air is approximately 1 m. The beta contributions from nuclides with maximum beta 

energies ≤ 350 keV were ignored and the beta dose rate at 1 m was set to 0.

Beta Dose Rates 

The publication of Cross et al. (1982) presents values of dose rates in air for a number of 

radionuclides of interest at NTS. The dose rate values (mGy • cm2 • MBq -1 • hr-1) are given at various 

distances, R, from the source out to the beta range in air. Where a value of R = 100 is presented, the 

dose rate at that distance, divided by R2 = 104 is used. In some cases dose rates are given for 

distances <100 cm and >100 cm. In those cases, the dose rate at 100 cm was determined by 

interpolation using the values for distances that span 100 cm. The interpolation method that provides 

the best fit depends on the beta spectral characteristics of the particular nuclide. However, for most 

nuclides, a linear interpolation yields the best fit, and, considering the overall errors involved, a linear 

interpolation may be considered adequate.

As an example, in the case of 22Na, dose rates are given at 97.5 cm and 105 cm. The dose rate at 

100 cm is determined as follows:

R2

J’(100) = R2J’(d1) - [(100 – d1) • (R2J’(d1) – R2J’(d3))] ÷ (d3 – d1) (C-3) 

R2

J’(100) = R2J’(97.5) - [(100 – 97.5) • ( R2J’(97.5) – R2J’(105))] ÷ (d3 – d1) 

R2

J’(100) = 9.7 - [(2.5) * ( 9.7 – 6.0)] ÷ (7.5)

R2

J’(100) = 8.5 mGy • cm2 •MBq-1 • hr-1

D(100) = R2

J’(100) ÷ 1002 = 8.5 • 10-4 mGy/hr per MBq

There are several radionuclides appearing in Table 2-2 and Table C-3 for which dose rate values 

have not included in Cross et al. (1982). In those cases, the following approximation (IAEA 1979) has 

been used:

D(100) = 0.008 mGy • hr-1 • n • C • d-2, (C-4) 

Where n = fraction of disintegrations that emit a beta

C = activity in becquerels

D = distance in meters

Since betas with energies ≥ 350 keV are absorbed in air, the value of D(100) is further reduced by the 

fraction, F, emitted betas that are so absorbed. As an approximation, F is determined by:

Fβ = (Eβ - 350)/Eβ, (C-5) 

ATTACHMENT C

BETA-PHOTON RATIO ESTIMATES FOR NTS WORK AREAS AND OPERATIONS

Page 11 of 14
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 114 of 121

where Eβ is the beta energy in keV.

For each radionuclide emitting betas with n different maximum energies, 

F = Σ (Eβi

 – 350)/Eβi, i = 1 to n (C-6) 

The nuclides in Table C-1 for which 1-m beta dose rates have been estimated using this 

approximation are indicated in bold.

C.5 SUMMARY OF BETA-PHOTON RATIOS

A summary and comparison of beta-photon ratios for the radionuclide inventory shown in Table 2-2 is 

listed in Table C-4. This illustrates the sensitivity of the ratios to the assumptions made in selecting 

the exposure geometry.

Table C-4. Geometry-dependent beta-photon ratios for NTS radionuclides 

(Table 2-2). 

Radionuclide 

of concern

Beta-photon ratio – Sv/Sv

Surface contamination Immersion Point source at 1 m

Ac-227 0.0E+00 2.5E-01 0.0E+00

Am-241 4.0E-04 1.2E-03 0.0E+00

Am-243 9.6E-06 7.9E-04 0.0E+00

AmBe-241 4.0E-04 1.2E-03 N/A

Ba-133 5.9E-05 2.4E-02 0.0E+00

Ba-140 7.8E+00 1.5E+00 2.7E+02

Cd-109 0.0E+00 0.0E+00 0.0E+00

Ce-139 0.0E+00 8.8E-02 0.0E+00

Ce-141 4.7E-01 1.5E+00 1.5E+01

Ce-143 1.0E+01 1.6E+00 5.6E+01

Ce-144a 1.4E+02 1.9E+01 1.5E+03

Cf-252 0.0E+00 3.6E-05 2.4E+00

Cm-244 0.0E+00 0.0E+00 0.0E+00

Co-57 5.0E-05 1.5E-02 0.0E+00

Co-60 3.7E-03 1.8E-02 9.8E+01

Cs-134 2.0E-01 8.9E-02 0.0E+00

Cs-137a 2.1E+02 3.9E+02 5.2E+00

Eu-152 3.3E-01 6.9E-02

Eu-154 1.1E+00 1.9E-01 1.5E+01

Eu-155 0.0E+00 1.1E-01 1.4E+01

Fe-59 1.5E-02 5.9E-02 0.0E+00

H-3 - - 2.8E-01

Hg-203 3.7E-06 1.9E-01 N/A

I-131 4.3E-01 4.1E-01 0.0E+00

I-132 1.9E+00 2.4E-01

I-133 5.4E+00 7.2E-01 1.8E+01

I-135 1.8E+00 2.4E-01

Ir-192 2.6E-01 2.2E-01

ATTACHMENT C

BETA-PHOTON RATIO ESTIMATES FOR NTS WORK AREAS AND OPERATIONS

Page 12 of 14
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 115 of 121

Radionuclide 

of concern

Beta-photon ratio – Sv/Sv

Surface contamination Immersion Point source at 1 m

Kr-85 2.2E+02 9.4E+01 7.5E+00

Lu-174 6.3E-03 1.8E-03 8.3E+03

Mn-54 3.1E-03 2.8E-04 1.8E+02

Mo-99 2.0E+01 2.7E+00 2.0E+01

Na-22 4.6E-02 7.5E-02 2.3E+00

Na-24 1.5E+00 1.4E-01 1.7E+01

Ni-63 N/A N/A N/A

Np-237 0.0E+00 5.2E-02 0.0E+00

Pm-147 0.0E+00 6.6E+02 0.0E+00

Pu-238 8.8E-04 1.2E-02 0.0E+00

Pu-239 3.2E-04 3.5E-02 0.0E+00

Pu-240 1.2E-06 1.2E-02 0.0E+00

Pu-241 0.0E+00 1.2E-01 0.0E+00

Pu-242 0.0E+00 1.5E-02 0.0E+00

PuBe-238 8.8E-04 1.2E-02 N/A

Ra-226 7.9E-05 3.1E-01 -

RaBe-226 7.9E-05 3.1E-01 N/A

Rh-100 2.4E-01 2.7E-02 9.9E+00

Rh-101 5.2E-05 4.2E-02 2.0E+00

Rh-102m 2.8E+00 3.7E-01 2.6E+00

Rh-103m 0.0E+00 0.0E+00 0.0E+00

Rh-106 5.5E+01 8.2E+00 2.2E+02

Ru-103 1.1E-01 6.4E-02 1.3E+00

Ru-106a 5.5E+01 8.2E+00 2.2E+02

Sb-124 1.6E+00 2.2E-01 1.4E+01

Sb-125 1.3E-01 1.2E-01 5.8E+00

Sn-113 0.0E+00 1.2E-02 7.4E+01

Sr-85 5.2E-02 7.7E-03 1.6E+01

Sr-89 1.8E+03 3.8E+02 3.8E+05

Sr-90 2.1E+02 8.4E+02 -

Ta-182 1.3E-01 8.2E-02 3.2E+00

Tc-99m 0.0E+00 4.4E-02 0.0E+00

Te-132 0.0E+00 1.2E-01 0.0E+00

Th-228 0.0E+00 1.4E-01 0.0E+00

Th-230 0.0E+00 6.6E-02 0.0E+00

Th-232 0.0E+00 9.0E-02 0.0E+00

Th-234 0.0E+00 7.1E-01 0.0E+00

U-233 2.6E-05 2.0E-01 0.0E+00

U-234 1.1E-04 7.5E-02 0.0E+00

U-235 3.6E-05 3.0E-02 0.0E+00

U-235 chaina 4.1E-01 2.0E-01 7.5E-04

U-238 0.0E+00 0.0E+00 0.0E+00

U-238 chainb 1.7E+02 3.1E+01 1.4E+02

Xe-133 1.1E-07 1.3E+00 0.0E+00

ATTACHMENT C

BETA-PHOTON RATIO ESTIMATES FOR NTS WORK AREAS AND OPERATIONS

Page 13 of 14
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 116 of 121

Radionuclide 

of concern

Beta-photon ratio–Sv/Sv

Surface contamination Immersion Point source at 1 m

Xe-133m 0.0E+00 4.5E+00 0.0E+00

Xe-135 6.2E+00 1.3E+00 9.4E+01

Y-88 3.6E-03 5.2E-04 0.0E+00

Y-90 1.3E+03 2.7E+02 1.5E+04

Y-91 8.8E+02 1.3E+02

Yb-169 6.0E-05 1.0E-01 2.5E-01

Zr-95c 2.7E-02 5.0E-02

Zr-97c 2.3E+01 2.6E+00

a. In equilibrium with Th-231. mSv/hr per MBq U = 235.

b. In equilibrium with Th-234, Pa-234m, and U-234. mSv/hr per MBq U = 238.

c. In equilibrium with daughter.

ATTACHMENT C

BETA-PHOTON RATIO ESTIMATES FOR NTS WORK AREAS AND OPERATIONS

Page 14 of 14
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 117 of 121

TABLE OF CONTENTS 

SECTION TITLE

D.1 Introduction ..........................................................................................................................

PAGE

117

D.2 Neutron Dose as a Function of Distance.............................................................................. 117

D.1 INTRODUCTION

One hundred atmospheric nuclear tests were conducted at NTS from January 27, 1951 to July 17, 

1962 (DOE 2000). Essentially all the neutrons accompanying a nuclear explosion are released either 

in the fission or fusion process. All of the neutrons from the fusion and more than 99% of the fission 

neutrons are produced almost immediately, within less than a millionth of a second of the initiation of 

the explosion (Glasstone and Dolan 1977). These are referred to as the "prompt" neutrons. In 

addition, somewhat less than 1% of the fission neutrons, called the "delayed" neutrons, are emitted 

subsequently. The majority of these delayed neutrons are released within the first minute, and so 

constitute part of the initial nuclear radiation. While the shielding afforded by even a few feet of soil 

attenuates the neutrons from underground tests by several orders of magnitude, only distance and 

minor attenuation by air contribute to the reduction of neutron dose from atmospheric tests. 

Therefore, in principle, significant occupational neutron exposure from an atmospheric test was 

possible if the person was close enough to the detonation point.

D.2 NEUTRON DOSE AS A FUNCTION OF DISTANCE

The spectra of neutrons produced by fission and thermonuclear weapons are illustrated in Figure D-1 

(Glasstone and Dolan 1977). The presentation is in terms of total neutrons per energy group per 

kiloton (kT) weapon yield. Based on these spectra, the neutron yield for a fission weapon would be 

about 7.7 × 1022 neutrons per kT, and for a thermonuclear weapon, the yield would be about 1.4 × 

1023 neutrons per kT. The fission neutron spectrum does not change very much to a distance of at 

least 1,600 yd. In contrast, for thermonuclear spectra, the relative contribution for neutrons above 

about 8-MeV decreases markedly for 400 to 1600 yd (Glasstone and Dolan 1977).

Neutron dose as a function of slant range can be determined from the neutron spectra as a function of 

distance (Figure D-2; Glasstone and Dolan 1977) for (1) fission and (2) thermonuclear air bursts. In 

the case of a fission weapon detonation, two curves are presented – offensive and defensive. The 

difference is the result of differences in weapons parameters. The defensive curves were used in 

prediction of slant range doses. The neutron dose as a function of slant range from Figure D-2 is 

compared in Figure D-3, which shows little difference between fission and thermonuclear weapon 

doses per kiloton. Using a radiation weighting factor of 20, the dose equivalent as a function of slant 

range is presented in Figure D-4 for a range of thermonuclear yields from 1 kT to 1 MT. Although the 

yield range is 3 orders of magnitude, the predicted 1-mrem dose equivalent slant range varies from 

about 3.8 km for 1 kT to 5.2 km for 1 MT.

ATTACHMENT D

NEUTRON DOSES FROM ATMOSPHERIC NUCLEAR TESTS

Page 1 of 5
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 118 of 121

Figure D-1. Neutron spectra for (a) fission weapon and (b) thermonuclear weapon 

per kiloton total energy yield (Glasstone and Dolan 1977).

Figure D-2. Initial neutron dose per kT as a function of slant range from (a) fission 

and (b) thermonuclear air bursts, based on 0.9 normal sea-level air density

(Glasstone and Dolan 1977).

a b

a b

ATTACHMENT D

NEUTRON DOSES FROM ATMOSPHERIC NUCLEAR TESTS

Page 2 of 5
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 119 of 121

Figure D-3. Comparison of the initial neutron dose per kiloton as a function of slant 

range from fission and thermonuclear air bursts (after Glasstone and Dolan 1977).

Figure D-4. Neutron dose equivalent as a function of slant range for atmospheric 

detonations with a range of weapon yields. 

The Defense Nuclear Agency has published a report on neutron exposure to DOD personnel during 

nuclear tests (DNA 1985). The results are based on radiation transport calculations using the ATR 

(Air Transport of Radiation) computer code. 

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

1.E+01

1.E+02

1.E+03

1.E+04

1.E+05

0 500 1000 1500 2000 2500 3000 3500 4000 4500

Range - Yards

Rads/kT

Fission Fusion 

1.E-03

1.E-02

1.E-01

1.E+00

1.E+01

1.E+02

1.E+03

0.000 1.000 2.000 3.000 4.000 5.000 6.000

Range - km

Rem

1 kT

2 kT

5 kT

10 kT

20 kT

50 kT

100 kT

200 kT

500 kT

1 MT

ATTACHMENT D

NEUTRON DOSES FROM ATMOSPHERIC NUCLEAR TESTS

Page 3 of 5
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 120 of 121

Calculations were made for NTS and the Pacific Proving Ground tests, taking into account differences 

in test conditions such as altitude (air density) and humidity. The results of these calculations are 

presented in terms of dose equivalent using a quality factor of 13.

The isodose contours for the continental detonations are shown in Figure D-5 (DNA 1985). Two tests 

were used to bound the range of possible results:

• Operation RANGER, Test EASY: The neutron spectrum was that of a pure fission device, 

typical of earlier weapons. The mean air density between the surface and burst height 

(330 m) is 1.18 × 10-3 g/cm3

, a very high value for NTS tests. The higher density and resulting 

increase in attenuation results in a lower neutron dose at a given range.

• Operation PLUMBOB, Test HOOD: The neutron source spectrum was that of a boosted 

fission device, typical of later weapons. The mean air density between the surface and burst 

height (460 m) is 0.992 × 10-3 g/cm3

, the lowest of any low-altitude NTS detonation. The low 

air density and enhanced neutron output result in a higher neutron dose at a given range.

Using the data in Figure D-5, the 1-mrem slant range at the surface for 1- to 80-kT detonations from 

about 4.1 km to 5 km are consistent with the data from Glasstone (Glasstone and Dolan 1977). 

Therefore, it can reasonably be concluded that personnel working on the surface beyond a range of 6 

km would receive neutron doses less than 1 mrem for the NTS atmospheric test series.

The isodose curves in Figure D-5 also provide information on neutron doses at altitudes as high as 

9 km (29,500 ft). This information is useful in assessing potential neutron exposures to air crews 

associated with the testing program.

ATTACHMENT D

NEUTRON DOSES FROM ATMOSPHERIC NUCLEAR TESTS

Page 4 of 5
Document No. ORAUT-TKBS-0008-6 Revision No. 01 Effective Date: 07/30/2007 Page 121 of 121

Figure D-5. 1-mrem neutron isodose curves for continental detonations (DNA 1985).

ATTACHMENT D

NEUTRON DOSES FROM ATMOSPHERIC NUCLEAR TESTS

Page 5 of 5